ML20101N368

Forwards Info Requested in NRC 951201 RAI Re IPE of External Event,Per GL 87-20,Suppl 4

ML20101N368
Person / Time
Site:	Sequoyah
Issue date:	03/29/1996
From:	Shell R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-87-20, TAC-M83674, TAC-M83675, NUDOCS 9604080356
Download: ML20101N368 (447)

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s s-YA Tennessee Valley Authonty. Post Office Box 2000. Sodcty-Da:sy. Tennessee 37379 March 29,1996 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In tM Matter of '

) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT (SON)- REQUEST FOR ADDITIONAL INFORMATION -

GENERIC LETTER (GL) 88-20, SUPPLEMENT NO. 4, "lNDIVIDUAL PLANT EXAMINATIONS OF EXTERNAL EVENTS (IPEEE) FOR SEVERE ACCIDENT VULNERABILTIES - 10 CFR 50.54(f)"

References:

1. TVA letter to NRC dated June 29,1995, "Sequoyah Nuclear Plant (SON) - Gbneric Letter (GL) 88-20, Supplement No. 4, ' Individual Plant Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)'"

2. NRC letter to TVA dated December 1,1995, " Request for Additional Information - Individual Plant Examination of External Events -

Sequoyah Nuclear Plant Units 1 and 2 (TAC Nos. M83674 and M83675)."

The purpose of this letter is to provide NRC with the information requested in Reference 2 above. TVA's response is provided in the enclosure to this letter.

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U.S. Nuclear Regulatory Commission Page 2 l

March 29,1996 i Please direct questions concerning this issue to W. C. Ludwig at (423) 843-7460.

Sincerely, R. H. Shell Manager SON Site Licensing Sworn og+ nd subscri)9p befor me t's/) day of 'I I I OAO I J 1996 '

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/ /YfD Duf y Com sion Expires .

Enclosure cc (Enclosure):

Mr. D. E. LaBarge, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike >

Rockville, Maryland 20852-2739 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 l

Regional Administrator ,

U.S. Nuclear Regulatory Commission ,

Region ll l 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 .

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I SEQUOYAH NUCLEAR PLANT l i

REQUEST FOR ADDITIONAL INFORMATION -

INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS - l l

'3EQUOYAH NUCLEAR PLANT l

UNITS 1 AND 2 (TAC NOS. M83674 AND M83675) l (B39 960326 001) l l

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Part I, Seismic Question 1 Question NUREG-1407, page 11, states that the seismic margin evaluation should utilize the NUREGICR-0098 median rock or soil spectrum anchored at 0.3g or 0.5g, depending on l the g-level and the primary condition at the site. It further states that the ground motion should be considered at the surface in the free field. Additionally, if secondary conditions such as shallow soil conditions are being considered, appropriate procedures j should be used to determine the free-field motion in the vicinity of those affected  ;

structures, and the capacity evaluation should take into account the effects of I soil-structure interaction, inasmuch as SON is a predominately rock site, with most I safety-related structures founded on rock, it should have used 0.3g specified at the rock outcrop instead of at the top of soil. The method used, which resulted in 0.3g at the soil surface, effectively reanalyzed the plant at the SSE grouad motion, which is inconsistent with the intent of the IPEEE program. Discuss the impact on results and insights if the analysis were performed in accordance with NUREG-1407 (i.e.,0.3g at the rock outcrop).  ;

Response

a. Background - The SON IPEEE program meets the intent of both NUREG-1407 and EPRI-6041, per our commitmentletter of December 23,1991. The discussion below provides additional detail regarding the objectives of the Seismic Margins Assessment (SMA),the process used to define the Seismic Margins Earthquake (SME) required by EPRI-6041, and the decision process related to the use of upgraded building models and analytical techniques; all of which taken together constitute the increased level of seismic demand that is necessary to apply the SMA methodology. The discussion below demonstrates that the Review Level Earthquake (RLE) chosen for SON is a consistent interpretation of the guidance in EPRI-6041. Additionally, a peer review approved the specification of the RLE and the definition of control motion. When the specification of the RLE is combined with the analysis methods used to generate seismic demand it will be seen that the SON IPEEE program complies with the intent of NUREG-1407.

b. Seismic Margins Earthquake - Section 3.2.2 of NUREG-1407 is clear that "the seismic margins evaluation should utilize the NUREG/CR-0098 media rock or soil spectrum anchored at 0.3 g or 0.5 g depending on the g level and primary condition at the site" and that " the ground motion should be considered at the surface in the free field (emphasis added)". As noted in the SON IPEEE report the site is overlaid with approximately 25-50 feet of soil. Accordingly, the RLE has been defined consistent with the quoted guidance (i e., a free-field top of soil peak ground acceleration (PGA) of 0.3 g).

c. Control Motion - Since some vital features (Diesel Generator [ DIG] Building, East Steam Valve Room, and the Refueling Water Storage Tank) are founded on soil, TVA was concerned that the application of the NUREG/CR-0098 spectral shape at the " ground surface" was unnecessarily conservative and technically inappropriate since it implies that top of ground motion is very broadband in energy content.

The analysis of actual earthquakes shows that top of ground motions usually display a narrow band shape dominated by the dynamic characteristics of the soil 1 l

deposit. Broadband spectral shapes (such as NUREG-0098 and R.G.1.60) are generally developed to represent a variety of causative earthquake effects and site conditions. Broadband shapes overstate the seismic threat for any one site because the site condition is no longer a variable. Therefore, the use of the NUREG/CR-0098 shape at rock is more consistent with the energy content experienced by rock-supported structures. This distinction is evident in Figures 3.1.3-1 (page 7 of 30) and 3.1.3-2 (page 8 of 30) from the SON IPEEE Final Report in that the response spectra for the free-field motion exhibit characteristic peaks corresponding to soil amplification for depths of 25-feet and 50-feet, respectively.

The approach used to define the control motion is consistent with ine analysis of actual earthquake records and is similar in concept to information previously presented by Drs. J. J. Johnson and Alejandro Asfura of EQE (Reference 1).

Dr. Johnson has also reported (Reference 2) on the results of research in seismic hazard analysis regarding the conservatism of the design ground motion for selected nuclear sites with a Housner Spectrum licensing basis, such as Sequoyah.

Reference 2 presents a comparison of typical Housner versus current treatment for selected elements of seismic analysis. Reference 2 also recommends the specification of the control motion to be either on a free surface on rock or soil (plant grade or on an outcrop). Finally, Reference 2 notes significant technical concerns related to specifying the control motion at foundation level as suggested ,

in the RAl. Those concerns are: >

1. Ignores the physics of the problem and the source of data used in developing design ground response spectra.

2. Results in motions on the soil surface whose response spectra display peaks and valleys associated with frequencies of the embedmentlayer which are fictitious.

3. Results in soil surface motion with a peak ground acceleration (PGA) that is typically greater than that of the control motion -- by seismological definitions, the design earthquake is increased.

4. Penalizes partially embedded structures compared to surface founded structures.

As discussed in Sections 6.2.2 and 6.3 of the IPEEE Final Report, Dr. l. M. Idriss provided peer review of the control motion issue. The scope of work for Dr. Idriss' review included the issue of control motion definition and generation of instructure response spectra (reference 3, copy enclosed). The peer review by Dr. Idriss concurred with the interpretation provided above (Reference 4, copy enclosed). In addition, the specification of the RLE in the manner described above is consistent with the guidance provided in Section 2 of EPRI NP-6041 (Reference 5) for selecting a Seismic Margins Earthquake. The categorization of the RLE in the manner described above is also consistent with Section A.4.4 of NUREG-1407 Decause Table 3.1 does not include SON in the listing of plants which require special attention to shallow soil conditions.

d. Seismic Demand - When the SON IPEEE program was developed TVA concwed with the premise of the program that meaningfulinsights into severe accident 2

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behavior would be gained by systematically examining the plant for seismic events which imposed loads considerably in excess of those from design basis conditions. l in fact, at industry workshops presented by NUMARC (now NEI), considerable l emphasis was placed on the fact that the most significant insights into plant vulnerabilities result from the application of the SMA process itself and is not highly dependent on the absolute level of SME defined. It was emphasized that the j performance of the earthquake experience-based walkdowns was the most  ;

important element of the program and would likely identify simple modifications that could increase the ruggedness of the success paths. The objective established for the IPEEE, therefore, was to define the seismic demand phase of the SMA in a realistic and technically appropriate manner.

As noted above, the SON IPEEE program is consistent with the spirit and intent of j the IPEEE guidance documents in that the plant was challenged seismically to l determine vulnerabilities. That challenge for Sequoyah comes from the use of upgraded structural models, current analytical techniques, and the use of the l NUREG/CR-0098 spectral shape, which significantly exceeds the design basis l Housner shape in critical frequency ranges. Figure 3 compares the '

NUREG/CR-0098 spectral shape anchored at 0.2 g with the design basis Housner spectrum anchored at 0.18 g and an alternative Seismic Margins Earthquake defined by an EPRI Uniform Hazard Spectrum (UHS) for 104 Non Exceedance  :

Probability (NEP) and anchored at 0.3 g The EPRI UHS spectrum is presented to provide a realistic baseline for determining the conservatism associated with the use of the NUREG/CR-0098 spectral shapes. The EPRi NP-6041 report notes that the most important frequencies of interest for an SME are approximately from 2 to 8 HZ. It also notes that the procedures in reference 6 allow toe high frequency portion of the UHS curve to be reduced to reflect the fact that high frequency motion is not as damaging as once thought. The EPRI UHS curve shows that the degree of conservatism associated with the use of the NUREG/CR-0090 shape is  ;

extremely conservative in the critical frequency range of 2-8 HZ. The 2-8 hZ range '

is critical because the natural frequency of most equipment is in the 2 8 HZ range.

l SON used updated structural models and analysis techniques for those structures l which are similar to corresponding structures at Watts Bar. As a result, the saismic l demand for the IPEEE differs, both in level and type, from the design basis seismic demand. For the rock supported structures additional seismic demand beyond .

design basis has been generated due to anhancements in the updated structural j models. The most significant factors are: '

1. The use of 3-D models for the Auxiliary Control Building (ACB), Shield Building (SB), and the Interior Concrete Structure (ICS).

2. The consideration of torsional effects at the extreme cross-sectional points in the ACB and ICS responses. Design basis structural models considered the torsional response at the center of mass to be applicable at all points across the cross-section. The consideration of torsion at extreme points in the updated l models results in an increase in torsional accelerations for items located away  ;

from the mass center and a significant frequency shift associated with the j location of the response spectrum peak (s).

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3. The consideration of revised torsional behavior for the ICS due to updated torsion constants.

4. Significant increase in vertical response and verticalinstructure response spectra due to analytical methods and the use of the same RLE for the horizontal and vertical directions.

The seisniic demand for IPEEE in the soil-supported structures is often less than the design basis demand because the design basis methods were extremely conservative. Reference 2 discusses in more detail the conservatism present in the soil structure ireteraction (SSI) methods for Housner vintage plants. The methodology used for the IPEEE SSI analysis, while more realistic and less conservative than the design basis methodology,is itself more conservative than the most current methods for SSI that have been benchmarked against the data in ,

Reference 7, such as the usage of computer programs SASSI and CLASSI. I

e. Conclusion - The SON IPEEE program complies with the intent of NUREG-1407 and i EPRI NP-6041 in that the plant has been subjected to increased seismic demand j resulting from the exceedance of the design basis Housner spectrum by the  !

NUREG/CR-0098 spectral shape and from the use of updated structural models and analytical techniques. As noted in Section 2 (page 2-10) of EPRI NP-6041, j

" Weakest-link elements are only found if one or more elements do not pass the ,

review procedure at the selected SME level." The RHR Heat Exchanger was l identified as a low ruggedness item and' modified. The increased seismic demand  ;

used in the IPEEE program is sufficient to result in the identification of vulnerable - j items, increasing the seismic demand to higher levels would not change the ranking of items in Table 3.1.41 and Section 8.1 of the IPEEE Final Report.

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References

1. Johnson, Dr. J. J. and Asfura, Dr. A. P., " Soil- Structure Interaction (SSI):

Observations, Data, and Correlative Analysis," Proceedings of the NATO Advanced  !

Study Institute on Developments in Dynamic Soil-Structure Interaction, July 8-16, '

1992, Antayla, Turkey.

2. Johnson, Dr. J.J., " Addressing the issue of Housner Seismic Design Criteria,"  :

Presented at The Fourth Symposium on Current issues Related to Nuclear Power  !

Plant Structures, Equipment, and Piping, December 9-11,1992,069do, Florida. l

3. A letter from EQE to Dr'. l.M Idriss providing the scope for his peer review of the j control motion and probabilistic generation of instructure response spectra for the  ;

SON IPEEE Program. EQE Correspondence No. 52207-0-022JJJ-93-0327 with  ;

subject of " Seismic Control Motion Definition for Sequoyah Nuclear Plant Seismk i Margin Assessment" and dated July 6,1993. (Attachment 1) i

4. A letter from Dr.l.M.ldriss to EQE providing the results of his peer review of the control motion specification and the probabilistic generation of instructure response spectra. Dated January 28,1994, with subject of " Seismic Control Motion Definition for Sequoyah Nuclear Plant - Seismic margin Assessment."  ;

(Attachment 2)  !

5. EPRI Report NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Final Report, August 1991.

6. J.W. Reed, et. al., " Analysis of High Frequency Seismic Effects," Palo Alto, CA, l Electric Power Research Institute; 1991, to be published. l

7. EPRI Report TR-100463," Spatial Variation of Earthquake Ground Motion for  !

Application to Soil Structure Interaction."

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Figure 3.1.3-1 from SQN IPEEE Final Report it ,

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Legends Notes: -

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• Accelerations in g's g'{1ygN{3,h,I05' n ~~~~~

Input PGA = 0.19 g Top of Lay 6 EL 680,. TCP of soil PGA = 0.30 g Base Rock Nation _ ,__

l Figure 3.1.3-1: Rock outcrop motion, and base rock and top of soil motion at the ACS location PnS12o7.otWQMPe3Vrw 3-83 1

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Part I, Seismic Question 2 Question NUREG-1407 (Section 3.2.5.8, page 14) suggests that human actions are to be clearly j identified and that the seismic IPEEE assessment assure that they are low enough in I probability so that they do not compromise the seismic margins. Typically, some of the I more risk significant human actions for an ice condenser plant include (among others):

connect ERCW as an auxiliary feedwater source, feed and bleed, RCS cooldown and l depressurization following LOCA, reduce containment spray pump flow, room cooling i recovery, and establish cold leg recirculation. The statements that all paths are twofold j redundant, written procedures provide guidance, and technical specifications exist do )

not provide assurance that human actions can be accomplished after a seismic margin earthquake (SME) and that success paths will be available after an SME. Provide a list of human actions required for the safe shutdown path and the plant locations at which these actions will be performed.

Response

The following is a list of manual actions to be carried out in accordance with AOP-N05,

" Earthquake Abnormal Operating Procedure." The actions are arranged in order as that i are addressed in the AOP. It should be noted that human actions necessary to achieve safe plant shutdown as defined in Emergency Response Procedures are performed in seismically qualified structures.

PROCEDURE STEP # ACTION ACCESS TO ACTION LOCATION AOP-N.05 2.O[3a] Check 0-R-113, Seismic Auxiliary Instrumer" Room Alarm Panel (seismically qualifica building) l 2.0[4] IM's perform SI-657, All data collection locations retrieve Earthquake Records are in seismically qualified structures I

SI-657 6,1 Collect data 0-XR-52-75. Auxiliary Instrument Room Strong Motion Accelegraph (seismically qualified building) 6.2 Collect data 0-XR-52-77 D/G Building Triaxial Time-History (seismically qualified building)

Accejegraph 6.3 Collect data 0-XR-52-89, D/G Building 1 Response Spectrum Recorder (seismically qualified building) i

6.4 Collect data 0-XR-52-88 Auxiliary Control Room Response Spectrum Recorder (seismically qualified building) 6.5 Collect data 0-XR-52-87 Unit 1 Containment el 734 Response Spectrum Recorder (seismically qualified building) 6.6 Collect data 0-XR-52-86 Unit 1 Annulus el 680 Response Spectrum Recorder (seismically qualified building) 67 Collect data 0-XR-52-83 Unit 1 Containment el 707 Peak-Recording Accelerometer (seismically qualified building) 9

4 SI657 6.8 Collect data 0-XR-52-82 Unit 1 Containment el 702 Peak-Recording Accelerometer (seismically qualified building) 6.9 Collect data 0-XR-52 84 Main Control Room Peak-Recording Accelerometer (seismically qualified building) 6.10 Collect data 0-XS-52-79,80, Unit 1 Annulus el 680

& 81, Horizontal Triggers (seismically qualified building)

AOP-N.05 2.Ol6a] Monitor ERCW traveling ERCW Building screens (within three hours seismically qualified building) of event) 2.0[7] Perform App A., Deenergizing C & A Vent Boards in Control Heaters Building (seismically qualified building) 2.0[7] Shutdown per 0-GO-5 & 0-GO-6 Various locations as listed below 0-GO-5 5.3[10c] Align steam seals (not a Turbine Building required action for safe shutdown) 5.3[14a] Shutdown a Main Feed Pump Turbine Building (not a required action for a safe shutdown) 0-GO-6 5.1[5f] Open MSR steam inlet valves Turbine Building (not a required action for a safe shutdown) 5.2[2c] Remove heaters from service Turbine Building (not a required action for a safe shutdown) 5.2[2d] Open extraction line drains Turbine Building (not a required action for a i safe shutdown) I 1

5.3113a1 IMs perform Auxiliary Building el 759 1,2-PI-IFT-99-OP4.0 (seismically qualified building)

Verify P-4 contacts 5.3[16] Deenergize Main Feed Pump Control Building el 669 trip bus (not a required (seismically qualified building) action for a safe shutdown) 5.3[19] Deenergize Main Turbine Control Building el 669 trip bus (not a required (seismically qualified building) action for a safe shutdown)

AOP-N.05 2.Ol121 Stop Control and Service Air Turbine Building Compressors (not a required action for a safe shutdown) 10

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l Part I, Seismic Question 3 j i

Question The submittallists walkdown anomalies (pages 3-7 through 3-11) which are a compilation of " unsatisfied caveats"in the screening and evaluation worksheets (SEWS). The categories of these anomalies include: potential functional failures, inadequate anchorage and load path, Category 11 to I proximity issues, and inadequate clearance. The submittal noted these items were addressed by means of maintenance or work requests. It is not clear, however,if the equipment capacity assessments used the as-found condition, or the prospective fixed condition, when screening out and i' evaluating equipment. Describe how the walkdown, screening, and seismic margins assessment accounted for the anomalous conditions of the equipment listed on  :

pages 3-7 through 3-11 of the submittal.

Response

The seismic margins assessment accounted for the anomalous condition of the equipment in three ways.

First, for the majority of equipment, the assessment uses the as-found condition.

Second, for one item of equipment (as discussed in the letter transmitting the final IPEEE Report), the assessment uses the " prospective fixed condition" (although TVA did not complete installation of the " prospective fixed condition" before issue of the final report). Problem Evaluation Report (PER) SO950426PERdocuments this deficiency. The deficiency is that the clearance between a reactor coolant pump seal return valve and the crane wall was not sufficient to prevent contact after system heatup. TVA plans to correct the deficiency during the U2C7 outage (Work Request C341196).

Finally, for the remaining equipment, the assessment uses the " prospective fixed condition" (because TVA completed installation of the " prospective fixed condition

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before issue of the final report).

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1 Part I, Seismic Question 4 Ouestion 4 The submittal states that a relay may be considered nonessential (i.e., pre-screened) if it does not prevent success of the associated desired function and does not cause the occurrence of an undesired event. The example is cited that battery chargers and associated relays are not essential because ample residual capacity is available to start diesels. However, once diesels are operating, battery chargers are essential to maintaining DC power for the full 72-hour success criterion. Provide an assessment of the need for battery chargers for the success paths over 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and, if found to be needed, provide the HCLPF of battery chargers and an assessment of the adequacy of the associated relays.

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Response 4 The capability of the D/G battery to start a D/G during design basis events such as loss of off site power (LOOP) and LOOP concurrent with an accident has been evaluated by the D/G system supplier, MKW Power Systems. MKW maintains the design basis for the capability of D/G to start within the required times for these events. Except for the first few seconds in the duty cycle, the battery, with the charger not operable, is capable of supplying all the loads without dropping below the minimum battery voltage

of 105 volts for 30 minutes when the battery is at the lowest expected temperature of 60 F and at the "end-of-life" condition (80% capacity). Further, TVA has performed an independent analysis of the D/G battery that supports the conclusions provided by MKW. The D/G battery, due to motor transients, will momentarily drop below 105 volts during the first 10 seconds of the duty cycle. However, all components will have adequate voltage to support the safety function.

A normal start of the D/G would be completed in approximately 10 seconds. This would involve removal of less than 0.2 amp-hour from the D/G battery. After this period, the battery will be supplying normal control D/G DC load current of approximately 1.19 amps. The D/G battery configuration at SON consists of 57 C & D type 3DCU-9 Lead Calcium cells (19-3 cell units,4 positive plates per cell) with rated capacity of 88 amp-hours at the 8 Hr. rating to 1.85 F.V. per cell at 77oF. Derating the battery for aging (80% capacity) and temperature (60 F) provides an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rating of 63 ampere-hour capacity. This would conservatively provide for approximately 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> of service at the 1.19 amp normal D/G control circuitry loading value.

This would be ample time to repair the damaged charger, if possible, or to provide an appropriate emergency 125V DC charger or power supply to continue operation of the D/G for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This would be accomplished even on loss of the D/G battery system

. problem annunciators. Post event walkdowns would identify D/G battery charger I damage and/or inoperable condition in a time to implement charger repair or substitution prior to the D/G becoming inoperable.

Part I, Seismic Question 5  ;

Question Page 4-79 of the submittal states that fires in the turbine building indirectly affect auxiliary feedwater. The auxiliary feedwater system is one of the safe shutdown systems, yet the effect of turbine building structural or masonry wall seismic failures on auxiliary feedwater (or any other system) were not evident from the submittal. Provide an explanation of how turbine building failures could affect the auxiliary feedwater system. If the effect could be significant, then provide the turbine building failure modes, locations, and screening results or HCLPFs.

Response

The AFW system is designed to function as an engineered safety feature. Suitable l piping, components, and trained power supplies are provided to assure that the system safety function (i.e. supply sufficient feedwater flow to the steam generators to remove primary system heat in the event of a loss of MFW flow)is met (assuming the worst j case single failure). The system interfaces with both the condensate system and the

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T essential raw cooling water system. To ensure the best possible steam generator water chemistry, the AFW system is usually aligned to the condensate system for AFW pump suction. A portion of the AFW pump suction piping from the condensate storage tanks is non-safety grade. This piping extends from the turbine building wall through the auxiliary building to check valves upstream of the AFW pumps. This nonsafety grade source of auxiliary feedwateris backed up by a safety grade source from the essential raw cooling water system. While the essential raw cooling water does not meet the same chemistry requirements as the condensate system, adequate flow will be provided to ensure that the AFW system will meet system functional requirements. Any failures in the turbine building will only affect the non-safety portion of the AFW system. The system will continuo to meet all safety functions as a result of the essential raw cooling water system interface. Further review of turbine building failure modes is not required.

Part I, Seismic Question 6 Question Provide walkdown notes and fragility calculations notes (as applicable) for the following components: ice condenser, auxiliary building roof diaphragm, ERCW MCC anchorage, lowest capacity masonry block wall affecting safe shutdown equipment list components, and RWST.

Response

Ice Condenser Walkdown Notes (do not exist)

Fragility Calculations

1. "lPEEE Seismic Margin Evaluation Containment and Ice Basket Seals", Calculation Number SCG-SM-0044 (B39 950609 027) (Attachment 3).

2. " Seismic Margin Assessment for The Ice Condenser Ice Basket Latice Frames",

Calculation Number SCG-5M-0047 (B39 950609 030) (Attachment 4).

3. " Seismic Margin Assessment for the Ice Condenser Top Deck Structure", i Calculation Number SCG-SM-0048 (B39 950609 031) (Attachment 5).

4. " Seismic Margin Assessment for The Ice Condensor Ice Basket Lower Support Structure", Calculation Number SCG-5M-0049 (B39 950609 032) (Attachment 6).

Auxiliary Buildina Roof Diaohraam Walkdown Notes (do not exist)

Fragility Calculation (excerpts enclosed in Attachment 7)

"lPEEE Seismic Margins Evaluation Auxiliary Control Building",

Calculation Number SCG-5M-0046 (B39 95 0609 029).

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ERCW MCC Anchorace Walkdown Notes (Attachrnent 8)

Fragility Calculation l

" Seismic Margin Assessment for the 480 V ERCW MCC', Calculation Number '

SCG-5M-OO27 (B39 950609 011) (Attachment 8).

Lowest Capacity Masonary Block Wall Walkdown Notes (do not exist)

Fragility Calculation "lPEEE Seismic Margin Evaluation Masonary Walls", Calculation Number SCG-5M-0042 (B39 950609 025) (Attachment 9).

RWST Walkdown Notes (Attachment 10)

Fragility Calculation "lPEEE Seismic Margin Evaluation Refueling Water Storage Tank" Calculation Number SCG-SM-0041 (B39 95 0609 024) (Attachment 10).

l 1 \

i t,

l 14

l Part II, Fire Question 1 Question 1 The control room was screened out using the qualitative argument that it is  ;

continuously occupied, the transient combustible control program is effective,  !

combustibles are separated from ignition sources and restrained fron falling into )

cabinets, and unit shutdown is available from the alternate shutdown panel. This is )

considered to be an insufficient basis to screen out the control room. The ability to i uppress a fire and achieve safe shutdown from another location does not mean that it )

will certainly be done. For example, the argument ignores the potential effects of l cabinet-initiated fires, which are an important fire source in the control room, that might !

force control room evacuation. In all fires PRAs, in which the frequency of cabinet l fires, the potential for suppression, and the conditional probability of successful )

operator actions and remaining equipment for a realistic selection of scenarios are ,

analyzed in detail, the control room emerges as a significant contributor to the overall '

l core damage frequency.

l Describe the change in the screening status of the control room and the total estimated I core damage frequency of unscreened compartments owing to quantitative analysis of the following: (1) frequency of cabinet fires, (2) probability of control room evacuation, l

(3) the probability of non-suppression, and (4) the conditional probability of safe

shutdown, including hardware and operator actions within and outside of the control I l room. Identify and describe vulnerabilities that may emerge from these considerations.

Response

Sequoyah did not perform a quantitative analysis of the Control Room based on the following information: TVA control room operators are highly trained for remote shutdown panel operation from the auxiliary control rooms. The object of the IPEEE Fire Analysis is to identify vulnerabilities and insights from the most important areas of concem. The Sequoyah Control Building, which contains the Main Control Room (MCR), is separated from the Auxiliary Building and the auxiliary control rooms by a l three foot thick poured concrete wall rated as a three hour fire barrier. The MCR was screened from further review based on: (1) The control room is continuously occupied by two control room crews, (2) The transient control program limits the amount and location of combustibles allowed in the control room, (3) Combustibles are generally separated from ignition sources and are restrained from falling into control cabinets, and (4) Shutdown of either unit can be obtained from the auxiliary control rooms which are outside the MCR.

l The reliability of safe shutdown equipment operated from the auxiliary control room, by design, should not be affected by any fire in the MCR. The equipment conditional probability f ailure is not affected when operated from a safe shutdown panel. Specific l

procedures define actions required for control room abandonment and actions required in the auxiliary control rooms. The operators are well trained for remote shutdown panel operation, to know what safe shutdown equipment is available at all times, to know how to operate the equipment, and to use the remote shutdown procedures.

15 l

Based on this additional qualitative analysis, a quantitative analysis was not performed.

The importance of the MCR is recognized. There are no new insights or vulnerabilities expected to be gained by the requested quantitative analysis which would differ from other IPEEE findings.

Question 2 The relay room is separated from the cable spreading room in Sequoyah. Because the cable spreading room has no cabinets,it was understandably screened out. However, the fire frequency of the relay room (which contains cabinets) was also screened out on the basis of a fire initiator frequency being low in comparison with other similar plants and with what would be expected from the FIVE databases used in this study.

Furthermore, the manual suppression failure probability of 0.1 was justified on the basis that this room is continuously manned. It is not typical for a relay room to be continuously manned. Given the numerous potential scenarios that can occur owing to relay room fires, the conditional core damage frequency of approximately 1X10 appears low in comparison to comparable rooms, such as the 125V Battery Board Room.

Provide a description of analyses performed for the relay room that shows the following: (1) derivation of the fire frequency, with consideration of all fire sources in the compartment;(2) fraction of time the relay room is manned, the fire protection training of those occupying the room, the fire protection equipment available within the room, and the fire brigade response time for that room; and (3) derivation of the conditional core damage frequency that accounts for all sources and scenarios in the relay room, including the potential for abandonment and use of the Auxiliary Shutdown Panel.

Response 2 The relay rooms were analyzed in accordance with the FIVE Methodology and screened based on the FIVE Screening Methodology and reasonable engineering assumptions for the specific fire area. This room is located adjacent to the MCR, has a relatively low ignition source frequency, automatic detection, manual suppression by portable extinguisher and fire hose, a fire load of less than 10 minutes, and response to detection would be immediate. At the time of the IPEEE walkdowns, the relay room was continuously occupied, but it is generally unoccupied. Fire in any room adjacent to the MCR is recognized as having the potential to cause abandonment of the MCR and no further evaluation is required. The overall frequency of a fire occurring and damaging shutdown components (F3 ) in this room was 4.49E-08 which is significantly smaller than the enhanced approach used by TENERA which screened rooms at a more conservative 1E-7 instead of the 1E-6 allowed by the FIVE Methodology.

Question 3 The separation of Thermo-Lag wrapped cables into separate, virtual compartments,in l

effect, assumes that the wrap is effective in precluding fire damage of the cable from l

any fire in the original compartment. A fire in the original compartment, therefore,is treated as not being able to damage wrapped cables within it no matter where the

source fire is. This assumption is contrary to FIVE guidance which suggests that 1 l

hour fire barriers are not to define the boundaries of compartments. Therefore, this 16

. . . - = . . _

assumption is not considered to be an acceptable approach. Furthermore, the basis for the extremely low fire initiator frequency assigned to each wrapped cable (3.1 X 104 per year) was not stated in the submittal. Other studies allow damage to occur to these cables if a fire propagation analysis concludes that the cable damage criterion has been reached. The concern is that the assumption of isolation of Thermo-Lag wrapped cables coupled with the low assigned fire initiator frequency has understated the fire risk associated with compartments that contain these cables.

Describe the change in the list of screened and unscreened compartments, and the change in the estimated core damage frequency of unscreened compartments, obtained from consideration of an approach that considers fire propagation from nearby sources (e.g., transformers, transient combustibles) to wrapped cable. Identify and describe vulnerabilities that may emerge from this consideration.

Response 3 Each Thermo-Lag 330-1 wrapped cable was inspected during the IPEEE walkdowns as was fire loading. Although provided with detection and suppression, for rooms where area fire loading could result in failure of the wrap, the protected systems were failed for the calculation of the area core damage frequency in accordance with the FIVE Methodology. Since the protected systems were failed, there should be no valid concerns about understated fire risks for compartments that contain these cables. I Therefore, the request presented in paragraph 2 for this question is not applicable for l the Sequoyah analysis. '

Question 4 The FIVE methodology requires an estimate of the Critical Combustible Loading that would be necessary for a target / source combination to exceed the damage threshold of the target. If a Critical Combustible Loading is present, then a time to damage, to ,,, is calculated. The probability of non-suppression depends on the relationship between t o,,

and the time demonstrated for automatic and/or manual suppression. The Sequoyah fire study employed a simplification of the FIVE method in that it did not base its manual and automatic non-suppression probability estimates on the calculation of a Critical Combustible Loading. Instead,it assumed blanket screening values for non-suppression probability. The study claimed that the use of a 0.3 screening value for manual non-suppression in untended rooms is conservative. Plant fire drill data, shown in submittal Tables 4.9-1 and 4.9-2, indicated that 0.3 corresponds to the likelihood of the fire brigade failing to arrive at the scene within 10 minutes after detection. A true suppression time that includes detection and extinguishment is not provided. The value of 0.3, therefore, is clearly not conservative for those compartments (at least 30% of them) for which the total suppression time exceeds ten minutes. The treatment of automatic suppression also underestimates the potential for fire damage because it assumes that the detectors are well placed and detection time is zero. The automatic suppression unavailability used in the submittal assumes full compliance with NFPA standards. Such compliance has not been demonstrated for Sequoyah. The failure of automatic suppression, employed in the study, is simply the system unavailability.

Describe the change in the list of screened and unscreened compartments, and the change in the estimated core damage frequency of unscreened compartments, obtained from consideration of either: (1) use of compartment-by-compartmentsuppression 17

1 probabilities that take into account the competition between fire growth and manual and/or automatic suppression response times, or (2) use of a blanket nonsuppression probability that is clearly conservative for all compartments. Identify and describe vulnerabilities that may emerge from these considerations.

Response 4 Fire suppression and detection are sized and spaced to combat the postulated fire for j the given area and meets or exceeds the applicable Code of Record NFPA standards, I unless approved deviations apply. The suppression quantity in many locations significantly exceeds the minimum required quantity. The thermal / smoke detectors would not add a significant amount of time for response / extinguishment actions. Initial response times average approximately 6 minutes. The longest response time of 27 minutes was for a switchyard fire. Excluding this, the average response time drops from a 9 minute to an 8 minute average. Extinguishment times may actually decrease l where manual backup suppression is used promptly instead of automatic actuation ]

(fusing) of sprinkler heads. Backup suppression and compartmentation does not i significantly affect the required actions. Given the above information, the 0.3 value is  !

judged reasonable for the present conditions. Additionally, the enhanced approach screened rooms at 1E-7 instead of the 1E-6 allowed by the FIVE Methodology. The l frequency of the screening value could be increased based on additional years of fire suppression data, if available, but any gains would be probabilisticaly insignificant because of the 1E-7 screening value.

Part 11, Fire Question 5 Question 5 Table 4.9-2 (which appears to use Table 4.9-1)is not a valid analysis. First, the analysis does not use a sufficientlylarge or diverse sample of plant locations. Second, the nonsuppression probability reaches zero at 30 minutes (which is an insufficient duration). Third, the analysis does not include the detection and suppression times (after arrival). Describe the effect on compartment screening, CDF, and vulnerabilities using a statistical approach that does not have these shortcomings. Consider the modified approach when responding to Question (4).

I Response 5 The time required to suppress most plant fires at Sequoyah is considered minimal compared to the relatively fast response times. The data supplied is from actual drill response times. This data covers various floor elevations of several main structures at Sequoyah. .All available fire response times were used for the IPEEE analysis and data for additional years is not available. The thermal / smoke detectors used at Sequoyah respond quickly and would not add a significant contribution to the times for response and extinguishment. Recent security modifications permit an even f aster fire brigade response to more plant areas than was considered in the IPEEE. The longest response time of 27 minutes supports the Table 4.9-2 probability data for a 30-minute fire brigade response time.

l 18

Question 6 The initiating events and systemic or functional sequences identified for each fire source location in a compartment are crucial to the evaluation of conditional core

damage probability. The selection influences both the complement of equipment and the human actions that are assumed to be required to prevent core damage. A review of the reasonableness of the quantitative screening calculations in the Sequoyah fire IPEEE cannot be made because initiating events, accident sequences, a list of analytical assumptions, sources of uncertainties, the functional / systemic event trees associated with fire-initiated sequences, and human actions have not been provided in accordance with NUREG-1407, page C-4, items 9,10, and 11.

Provide the following for each unscreened compartment: (1) the initiating events analyzed, (2) the accident sequences and a word description of the accident sequences that does not rely upon knowledge of the top event identifiers in the event trees, (3) a list of key analytical assumptions used in the development of the conditional probability of core damage, (4) the functional or systemic event trees used in the fire analysis with a description of the top events, and (5) the key human actions of each sequence. Also, provide a general description of the sources of uncertainty in the fire IPEEE.

Response 6 The "Sequoyah Nuclear Plant Unit 1 Probabilistic Risk Assessment Individual Plant Examination" (IPE), dated September 1992, has previously been provided to the NRC in response to Generic Letter 88-20. The requested information and required review can be achieved by utilizing the applicable IPE information and the initiating events information contained in Section 3.5 and Appendix G of the Sequoyah IPEEE Phase l Report. The IPEEE initiating event information (Section 3.5 and Appendix G) was not provided in the IPEEE Report submittal of June 1995, but is being provided for this response (Attachment 11).

Question 7 The potential for fires to induce a small LOCA was not mentioned in the submittal. In particular, it is not clear how the potential for the fire-induced opening of pressurizer PORVs, or how seal LOCAs derived from loss of sealinjection and cooling, were treated, in addition, the submittal states that pressurizer PORVs were walked down to determine if a potential fire could affect them. However, the results or insights of this walkt own were not provided in the submittal. If the potential for fires to induce LOCAs was considered, provide a list of compartments and fire scenarios that included fire-induced LOCAs. If fire-induced LOCAs were not considered, describe the change in the list of screened and unscreened compartments, and the change in the estimated core damage of unscreened compartments, obtained from consideration of fire-induced LOCAS. Also provide the results of the pressurizer PORV walkdown.

Response 7 The IPE took into account system failures that could result in a seal LOCA. System failures identified in the fires analysis took these systems into account as appropriate.

The potential for getting a hot short without grounding out or without blowing the fuses is probabilisticaly insignificant and the cable routing of these valves only shows up in 19

I credible fire scenarios; therefore, the possibility of failing these cables due to a fire is ,

extremely small. Section 4.6, Plant Walkdowns, and Table 4.6.1, Walkdown Summary,  !

from the FIVE Final report was not provided in the IPEEE Report, but is being provided for this response (Attachment 11).

Question 8 The study used three methods to obtain the conditional probability of core damage given a fire. Two were based on the IPE event tree / fault tree models, and one was  !

based on a Risk Achievement Worth model. None of the methods were explained as  !

requested in NUREG-1407, Page C-4, Items 7 and 9. Describe each method, including l assumptions, and provide example application of each method for the following zones:

6.9 kV Shutdown Board Rooms 734-A02FS1 Heating and Ventilation 714-A03FS2 Relay Room 732-C13 Also include the following information for each of the above zones: (a) list of components and systems affected by the fire, and (b) other components and systems that are failed in the assessment of core damage.

Response 8 '

The IPE model was used to quantify the core damage associated with the systems  :

failed in a given fire scenario. The IPEEE Phase I report identified the systems and ,

components in each room and a description of the fire scenarios is provided in Table 4.5-1 of the IPEEE Report. Sections 3.0 - 3.4 and Appendix C of the Phase i Report was not included in the IPEEE Report submittal of June 1995, but is being provided for '

this response (Attachment 11). Portions of text from Section 3.2, Safe' Shutdown Equipment Unavailability, was included in the IPEEE Report in Section 4.7, Analysis of Plant Systems, Sequences, and Plant Response. The complete Section 3.2 text addresses Risk Achievement Worth and is being provided for this response (Attachment 11).

Part II, Fire Question 9 Question 9 The effect of fires on failure of decay heat removal, per U3I A-45, does not appear to have been addressed in an analytical manner. For examMe, some IPEEEs have performed a risk assessment resulting in a core damage frequency for fire-induced loss of decay heat removal scenarios. Other IPEEEs have explicitly reviewed the IPE models for loss of decay heat removal to determine the effect of fire on decay heat removal availability. Provide an explanation and the analytical basis for the statement on page 4-91 that no significant findings were identified that impact the decay heat removal system at Sequoyah. The explanation should include the following: (1) the ,

compartments that contain decay heat removal systems, and their support systems, as I identified in Section 3.4.4 of the Sequoyah IPE, (2) the potential of fire-induced damage 20

e of docay heat removal systems in these compartments, (3) the scenarios (with descriptive text) in each of the identified compartments that include heat decay removal systems, and (4) the fraction of fire-induced core damage frequency associated with loss of decay heat removal equipment or a similar measure of the importance of the effect of fire on decay heat removal equipment.

Response 9 USl A 45 addressing decay heat removalis specifically addressed in Section 3.4.4 of the IPE. A specific risk assessment analysis of the impact of decay heat removal on core damage frequency was neither performed nor warranted for this issue. The impact of decay heat removal was evaluated based on the system importance which are embedded in the IPE analysis. The importance measures considered for decay heat removal contain sufficient capabilities to aid in interpretation and evaluation of the impact to core damage. These are discussed in Section 3.4.4 of the IPE. As noted, there were no significant vulnerabilities of the Sequoyah decay heat removal systems identified.

Additionally, there werr., no significant findings impacting the decay heat remeval systems at Sequoyah ijentified during the IPEEE fire analysis. This was basec on the inherent low level of importance of the decay heat removal systems from the evaluation and conclusions of the IPE findings. Additionalinformation from plant walkdowns was also utilized to evaluate potential vulnerabilities.

i Part II, Fire Question 10 Question 10 The study assumes that passive fire-barrier elements (e.g., walls, floors, ceilings, and penetration seals) are 100% reliable. Such an analysis is not valid unless the assumption is adequately justified and it can be demonstrated that there are no paths through the barrier for the spread of damage. Provide such justifications and demonstration for high-hazard fire areas, such as; the turbine building, D/G rooms, cable spreading rooms, switchgear rooms, and lube oil storage areas.

Response 10 The Sequoyah Fire Barrier Qualifications were discussed in Section 4.9.2 of the IPEEE Report. The Penetration Seal Program Assessment for Sequoyah was completed in 1993 as noted in the iPEEE Report. The Fire Analysis used the FIVE Methodology to screen boundaries and the boundary integrity for the buildings and rooms noted was examined / verified during the walkdowns. This included the three (3) hour fire rated barriers that separate major buildings at Sequoyah (i.e., Turbine, Control, Auxiliary, and Service) and also provide separation for the D/G rooms, cable spreading room, switchgear rooms, and the lube oil storage area (s).

Sequoyah utilizes the " Defense-in-Depth" philosophy to compensate for worst case room fire durations. Significant fire loads in these buildings are generally not concentrated near the 3-hour barriers. The individual fire area fire barriers are completely sealed and the majority are generally sealed to depths greater than required.

This provides a barrier rating that exceeds requirements in some configurations, 21

e I

• I although no additional credit is taken. Prevention, detection, prompt fire brigade response, manual and/or automatic suppression, and fire area compartmentation provide assurance that there are no paths through fire barriers to permit the spread of damage. The fire barrier qualifications in the IPEEE Report meet the attributes of an adequate fire protection program as defined in the Sandia Fire Risk Scoping Study Evaluation. I Part II, Fire Question 11 Question 11 The fire ccmpartment interaction analysis (FCIA) is based on the assumption that fire barriers are effective as rated. For active fire barriers (e.g., a normally open fire door that gets closed by fusible link), the failure probability can be significantly high. Provide a list of compartments with active fire barriers, a description of the active barriers, and a discussion regarding qualitative screening of these (and their adjacent) compartments.

Response 11 ,

1 As described in Section 4.9.2 of the iPEEE, fire doors are inspected on a regular basis in accordance with nationally recognized fire protection standards. Recent fire damper and penetration seal evaluations were also identified. Fire doors, dampers, and penetration seals have all recently been reevaluated and are regularly inspected in detail by TVA. For the IPEEE, the adequacy and acceptability of active fire barriers (i.e., roll up/ sliding fire doors and fire dampers) was documented on walkdown sheets for all compartments. The FIVE methodology only requires the plant to demonstrate that fire barriers and their components are being inspected and maintained on a regular basis in accordance with plant procedures and that appropriate compensatory measures are being taken when discrepancies are found. The fire barrier qualifications in the IPEEE Report meet the attributes of an adequate fire protection program as defined in the ,

Sandia Fire Risk Scoping Study Evaluation. l Part II, Fire Question 12 l

l Question 12 l The submittal states that some mercoid switches remain at the plant. One aspect of ,

the seismic-fire interaction issue is the potential for CO2 system controller low- I ruggedness mercoid slave relays to trip D/Gs or isolate room cooling. The submittal discussion did not address this aspect of the issue. Provide the location and identify the function of the mercoid relays and an assessment of the potentialinteractions caused by seismic activation of these mercoid relays.

Response 12 TVA purchased the CO2 Fire Protection system for the D/G Building on contract 83523 (refer to CCD NO:1-45N1699-1 RO). The contract includes the specification that " ....

Relays shall be dry snap-action type and suitable for seismic conditions .... There shall 22

.e 1

I be no mercury contacts on any electrical relays ...." The seismic qualification test report for the fire extinguishing equipment (Gaynes Engineering & Test Laboratory Report # 731031,807 89 0915 006) shows that the relays are rugged and will not inadvertently trip the D/Gs or isolate room cooling due to a seismic event.

23

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ATTACHMENT 1 L M.EdJP2mm P. O. Bear 330 Davis,CA 95027-0330 FAX b916) 758-1104 TEL' (916) 758-5739 \ 7S2-5403 l

January 28,1994 Dr. James J. Johnson President

EQE Engineering Consultants

! 44 Montgomery Street, Suite 3200 San Francisco, CA 94104

Dear Dr. Johnson:

l

Subject:

Seismic ControlMotion Definition

! for the Sequoyah Nuclear Plant Seismic Margin Assessment l

As requested in your letter dated July 6,1993, I have reviewed those aspects of the j seismic margin assessment related to definition of the seismic control motion and to l earthquake ground motions calculated at the site of the Sequoyah Nuclear Plant. This seismic margin assessment is being completed by EQE Engineering Consultants as part of the IPEEE for this plant.

My participation consisted of reviewing material supplied by EQE and attending a meeting at EQE's Offices on August 31,1993. At that time you and your staff outlined the procedure being used by EQE to develop the rock outcrop input motion to the soil profiles at the site, the approach for establishing best estimate as well as appropriate variations in the dynamic soil properties for these soil profiles and the methodology for evaluating the response of these soil profiles.

Based on the material reviewed and on the material presented at the meeting, the procedures, the material properties and hence the results of the ground response analyses appear reasonable.

It has been a pleasure to participate in this review with you and others at EQE.

Sincerely yours, ,

\

-r Ab_u ~

l r

( .

I. M. Idriss 1

FEB-13-19% 08:48 CORP ENGG t1&! / MN CHATT 615 751 8186 P.013 h ATTACHMENT 2 kbbNbNO 5 Engineering

@k a Safety ' Dup EQE Correspondence No. 52207 0-022 mw w. JJJ 93 0327 July 6,1993 Dr. I. M. Iddss P. O. Box 330 Davis, CA 95617-0330

Subject:

Seismic Control Motion Definition for Sequoyah Nuclear Plant Seismic Margin Assessment

Dear Dr. Idriss:

y As we have discussed on the phone EQE is performing a seismic margin assessment (3MA) for TVA's Sequoyah Nuclear Plant (SQN). As a part of this effon. EQE is committed to a peer review of this SMA.

We are currently developing the specification of the seismic demand in the Seismic .

Category I building structures in tenns ofIn-stnicture response spectra. We are developing the in structure spectra based on the guidance provided in NUREG 1407 and EPRI Report NP-6041. The attached paper documents,1) The specification of the ground control motion, and 2) The probabilistic generation ofin-structure response spectra.

EQE is committed to a peer review of the SQN SMA project. We would like your assistance and request that you perform the peer review o. the control motion specification and the probabilistic generation ofin structure response spectra as presented in the attached paper. As ! Indicated, I believe this review should not take '

more than 8 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of your time. Ms. Elaine M. Brovent will be forwarding the contractual items to ensure compensation for your participation. If you have any administrative questions, please feel free to contact her at the number below.

We look fonvard to your participation on this project and if you have any questions, please do not hesitate to call Dr. Jonna Arros or me at (415)989-2000.

Sincerely, EQE Engineering Concultants Dr ames J. Johnson President

/

JJJ:JKA:emb Attachments cc: Bob Enis (TVA) w/ attachments J. K. Arros (EQE w/ attachments T. R. Kipp (EQE) w/ attachments emb/tvakkts. doe

• 4 Montgomey SrNet. Suite J200. San franetwo. CA 94104 U%\ , Te:cphone (41.4) Wl9-2(XX) , TAN (41!) 4)).,4107

FEB-13-19% 08:49 COPP EN3G M&I / MH CHATT 615 751 8186 P.014 stacrLenent to EQE Correspondence No. 52207-O-022 I.0 INTRODUCTION As the first phase of the SQN IPEEE Seismic Margins Assessment (SMA), seismic demand in terms of in structure response spectra is developed for the following Seismic {

Category I building structures:

1. Reactor Building, consisting of:

Shield Building (SB) .

Steel Containment Vessel (SCV)

Interior Concrete Structure (ICS)

) 2. Auxiliary / Control Building (ACB)

3. Diesel Generator Building (DGB)

4. East Steam Valve Room (ESVR)

5. Emergency Raw Cooling Water Pumping Station (ERCWPS)

The methodology for the generation of'in-structure response spectra involves several elements:

Specification of control motion Development of structural models Development of soil models Modeling of soil-structure interaction Generation of in-structure response spectra using either:

a) Deterministic approach, or b) Probabilistic approach

/ The objective of seismic margin assessments is to " realistically" assess margins inherent in the designs of plant structures and components. Consequently, when determining the j seismic demand of structures and of components within the structures due to the review '

level earthquake (RLE), structural response is to be computed as " median centered",

FEB-13-1996 08:49 CORP ENGG M&I / MN CHATT 615 '751 8185 P.015 l EQE Correspondence No. 52207-0 022

, Page 2 of 9 l

eliminating the conservatisms intentionally introduced in the design response calculations.

To this end, seismic demand for the SQN SMA is developed using the probabilistic approach assuming the RLE as an 84% non-exceedance probability (NEP) event and treating structural damping and natural frequencies as well as soil material damping and shear moduli as random variables.

This report discusses the following specific aspects of the methodology for generating the in-structure response spectra for the SQN SMAi e

1. Specification of control motion,i.e., the ground motion spectral shape and the control point where the motion with the specified spectral shape is to be applied

2. The probabilistic method of generating in-structure response spectra for SQN SMA 2.0 SPECIFICATION OF CONTROL MOTION Soil conditions, including the depth and composition of the soil layers overlaying the l base rock, significantly affect the free field ground motion induced by a given seismic ' '

event and must be considered when defining the control motion. SQN site soil conditions are discussed in Section 2.1 and the control motion is specified in Section 2.2 2.1 Soil Conditions Natural soil condition at the SQN site, before construction, consisted of base rock overlaid with a soil layer varying in thickness from approximately 25 feet at tha ACB I location to more than 50 feet at the DGB location. Top of rock was at approximately 680 foot elevation at the ACB location and at 660 foot elevatiori at the DGB location.

Typically, the low strain shear wave velocity in the rock is approximately 5000 ft/see and in the overlaying soll within the range of 1100 to 2000 ft/sec.

At the RB and ACB locations, soil was excavated such that these buildings are founded ,

on rock with the finished grade at 705 foot elevation. The ESVR, located immediately adjacent to the RB,is founded on caissons that extend to rock. Soil parameter values in the area of the RB, ACB (for the excavated soil), and the ESVR are shown in Table 1.

I

FEB-13-1996 08:50 CORP ENGG M&I / MN CHATT 615 751 8186 P.016 bbb e pondence No. 52207 0 022 Page 3 of 9

'The DGB is founded on top of a 52.5 foot thick soil layer with the bottom of its foundation at 712.5 foot elevation. Soil parameter values at the DGB location are shown in Table 2.

The ERCWPS is founded on rock with the bottom portion of the structure, constructed of tremie concrete, immersed in water.

Based on the above descriptions, soil-structure interaction is a consideration only for the DGB and the ESVR.

2.2 Control Motion In accordance with the guidance of NUREG-1407 (Reference 4.1) and the specification of SQN as a 0.3 g RLE site, the control motion at the SQN site is based on the following:

1. Minimum free field PGA at top of soil is 0.3 g

2. Spectral shape is defined by specifying NUREG/CR-0098 median rock spectrum as the rock outcrop motion The level of the rock outcrop motion (with NUREG/CR-0098 median spectral shape for i rock sites) was determined such that a free field, top of soil PGA of 0.3 g was achieved '

at a location where the soil layer is shallowest and the soil amplification lowest (ACB and RB location). Consequently, at locations with thicker soll layer, e.g., at the DGB location, the free field, top of soil PGA exceeds 0.3 g.

Definition of the control motion in terms of the rock outcrop motion results in a consistent ground motion specification for all structures within the SQN site.

Furthermore, this control motion definition results in an amplified, narrow-band spectral shape at the top of soil which is physically reasonable since any base rock motion, even broad banded, would result in a peaked top of soil response due to resonant soil amplification.

The control motion specification described above is illustrated in Figure 1 by three spectra plots: 1) Rock outcrop motion specified as the NUREG/CR-0098 median rock shape,2) Base rock motion calculated from the rock outcrop motion by considering the effect of the overlaying soil, and 3) Top of soil motion. Computer code SHAKE was j used for the iterative computation required to account for the strain dependent soil  ;

properties (shear modulus and material damping).

Based on Figure 1, a rock outcrop motion with a ZPA of 0.19 g rems in a PGA of 0.3 g and peak spectral acceleration of 1.2 g at the top of soil.

FEB-13-1996 08:51 CORP ENGG M&I / MN CHATT 615 751 8186 P.017 EQE Conesponden, No. 52207-O-022 Page 4 of 9

' The base rock motion is specified as the input motion at the base of the dynamic models for the rock founded structures, RB, ACB, and the ERCWPS, when deriving the in-structure response spectra.

Spectra within the DGB are derived including soil-structure interaction effects. The foundation input motion is determined by propagating the base rock motion through the -

52.5 feet thick soillayer at the DGB location. Figure 2 shows three spectra plots: 1) Rock outcrop motion (same as in Figure 1),2) Base rock motion, and 3) Top of soil motion at the DGB location, with a PGA of 0.317 g and a peak of 1.65 g, which is used as the foundation input motion for the DGB. .

(As all components in the ESVR to be evaluated within the SMA are located at elevations less than 40 feet above the grade, use is made of the provision that allows the evaluation of these components to the free field spectrum amplified by the factor of 1.5. Should any of the components in the ESVR have a natural frequency less than 8 Hz, additional considerations will be required.)

3.0 PROBABILISTIC GENERATION OF SPECTRA ~

' Section 2 of NP-6041 (Reference 4.2) provides guidance for computation of structural response:

For the specified SME, the elastic computed response (SME demand) of structures and components mounted thereon should be defined at the 84%

non-exceedance probability (NEP).

To determine the 84% NEP response, a probabilistic method of generating in-structure response spectra is used which accounts for the uncertainty in both the ground motion description and in the structural parameter values. Thirty separate earthquake response ,

analyses will be performed for each building. Each analysis uses a different earthquake excitation and a different set of structural parameter values. (Each earthquake consists ,

of three statistically independent component time histories, one in each of the three i principal directions.)  !

For this purpose, an ensemble of time histories is generated such that their spectral ordinates are log-normally distributed with a coefficient of variation (COV) equal to 0.25, 1 and the 84% NEP value matched to the NUREG/CR-0098 median rock shape.

The uncertainty in structural and soil parameters will be defined in terms of coefficients of variation (COV) of the structural modal frequencies and damping ratios and COVs of the soil shear moduli and soll material damping.  :

i

1 FEB-13-1996 08:51 CORP ENGG M&I / MH CHATT 615 751 8186 P.018 I EQE Corresponder

. No. 52207 O-022 Page 5 of 9  ;

Uncertainty in the structural properties is accounted for by representing structural natural frequencies and damping ratios as log-normally distributed random variables with specified median and coefficient of variation (COV) values. Similarly for soil, shear ,

modulus and material damping are represented as log-normally distributed with specified medians and COVs.

A single COV value is associated with each of the random variables as follows:

1 Structural Properties:

COV(modal frequencies) = 0.25 COV(modal damping) = 0.35 Soil Properties:

COV(shear modulus) = 0.35 COV(material damping) = 0.50 The Latin Hybercube method is used to generate a set of random multipliers that are applied to the best estimate values to obtain the parame'.er values used with each of the 30 time histories.

As a result of the 30 time history analyses,30 response histories are obtained at each location of the building where the spectra are to be generated. For a given location, the l seismic demand is defined as the 84% NEP spectrum as determined from the ensemble of spectra.

i REFERENCES 4.1 NUREG 1407, " Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities" 4.2 EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Seismic Margin", Revision 1, August 1991

/

i

1 FEB-13-1996 08:52 CORP ENGG fit! / tti CHATT 615 751 8186 P.019 E@E Correspondene No. 52207-O-022 Page 6 of 9 TABLE 1 SEQUOYAH NUCLEAR PLANT - IPEEE SHEAR WAVE VELOCITIES IN VICINrrY OF REACTOR / AUXILIARY BUILDINGS Devation (ft) Soll Profile Shear Wave Velocity (fthec) 705 - 690 Residual Soils or 1,100 -

Terrace Deposits ,

690 - 680 Residual Soils, Shale, 1,600 @ 690 to

, Wea$ered Rock 2,000 @ 680 Below 680 Rock 5,000 -

Note: For backfill rrerials use a shear wave velocity of 1,000 Nsee.

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FEB-13-1996 08:52 CORP ENGG M&1 / T1N CHATT 615 751 8186 P.020 Attachment to EQE Corresponden No. 52207-O-022 Page7 Of 9 TABLE 2  ;

SEQUOYAH NUCLEAR PLANT IPEEE SHEAR WAVE VELOCITIES IN VICINITY OF DIESEL GENERATOR BUILDING Elevation (ft) Soll Profile Shear Wave Velodty (ft/sec) 721.5 690 Residual Soils or 1,100 Tensee Deposits 690 - 600 Residual Soils, Shale, 1,400 @ 690 Weathered Rock to 2,100 @ 660 Below 660 Rock 5.000 Note: For backfill maternis uma a shear wave velocity of 1,000 Maec.

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r ATTACHMENT 3 TVA 10697 W CALCUL/\I'CNC [7[ff Page 1 PLANT / UNIT TIRE IPEEE Seismic Margin Evaluation SON 1/2 containment and fra Rackets &A KEY NOUNS (Consult CCRIS LIST)

MsPARING ORGANIZ ATION Ne/EQE Seismic Margin, IPEEE Each time these calculations are issued, preparers must ensure that the original (RO) RIMS accession number BRANCH / PROJECT IDENTIFIERS daar. is filled in.

~

Rev (lor RIM'S use) RIMS accession number

_ _ uv- . . .

liWW V 5h04 n27 APPLICABLE DESIGN DOCUMENT (S) R1 .

N/A R2 SAR SECTION(S) UNID SYSTEM (S) R3 N/A N/A Rovision 0 R1 R2 R3 Safety.related? Yes No )(

EGN No ( indicate Not Applicable) Statement of Problem Pre od d[1//s/t./ ) [ Compute the seismic Margin Hi g h-Confi denc e-o f-l ow- P rob abi l i ty i

I Ch' k of-failure (HCLPF) acceleration capacity o containment and

• • Ice Baske accordance with M. A . #the conserva ive Deterministic Approved Failure Margin (CDFM) approach f$rqL bA2/ 00

• d/h.9 List,+1,mges added

) use term by this revision List ait pages de!oted

,, by this revision spac. List all pages changed

• *1 by this revision ,

These calculations contain unverified assumption (s) Calculation contains special requirements or that must be verilied later. Yes )( No ] iimiting conditions.

Yes ]No {

Abstract The cal u1 tion documents the seismic margin assessment of the Cont 'nppnt and Ice 4

$ Basket gn support of the SQN IPEEE. The containment and ice basket 4 Fe shown to I exhibit seismic capacity in excess of the Review Level Earthquake (RLE). i 1

Wr be.c_.cel is not a calcxt erN ow ye.- %e. ve r e_ m e n+s e-f #5 P 3./

bd IF hot l4tcd in he CCTilS ddebase . Tar is not a desip 64su cci tw la +M . ,

l Coure cT No : rv-to49ov cc.rs: Nco 89 o 2.3 5 o o 6 (vQ l Cyl 8'f- 2.o j S v R.. y #CO fr*l d 2-3 6 6 d ~/ (u 2 co,d comoh

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V O Microfilm and store calculations in RIMS Service Center.

Microfilrn and destroy. O s

2 ..~ J.6 m.m. - . J _ ,~ ~ m. CRCU- / dim SQ" 00l' 1A j-

Page 2/ic)

'I REVISION LOG L

sas ser-coff x IPEEE Seismic Margin Evaluation 500- " 0034 Weby IllI9' Containment and Ice Baskets Date d Revision DESCRIPTION OF REVISION Approved No.

0 Initial Issue The number of pages in this calculation is b .

Legibility evaluated and accepted for issue.

This applies for every page.

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JOB NO. ^' ~ o JOB 7 VM BY Nb) DATE '7 ' T -%

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w lik

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JOB NO. .C2ro7M JOB 7 v'A By d4) DATE 7- "'

CALC. NO. SUBJECT C m - '"' lec -

/ G & = 'r bc CHK'D M DATE 7 -l2.-M ,

M -5M-oo44'

! A eee.o usarc Caez.c. 19erece><s,,0xi av unieseir e-xx

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l A = 3&.4 3 ++' - sz ns.oz. E' I' - s e t. tz .io ' L .

f

[ 601./0" h r I,z 4L

• jd N \d : I? SC - 23I.4,2 322.I'2./O'l[ .

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1L T .zZ - 6 7 9. ? ? = A l, S S l'h Ta lHl&$. v .

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SHEET NO. I l'}

JOB NO. C?" ^ 7 JOB 7 V /' BY N/ DATE 'I l -M CALC. NO. SUBJECT Om 7-r 76-wr C-e L CHK'D DM DATE PMi

$CG GM ood4 Et . [R Thfto' /W [k-hl} hf/! [klC4 rn'fA fk thi

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% )

w SHEET NO. I'- li l JOB NO. @2W o'? JOB r v /1 BY Il/, ) DATE 7 ~ 7-M .

CALC. NO. SUBJECT S <' r '

'* 5 < /C' ' vr7 C a z- CHK'D DH DATE~Il-U

l. SCG SM &4 M- ioz,&oo kR ( th sc )

= (,2.31 '10 5 -lb c= 6 90 + l.17T 4 *).T -

900.??S T: h_fc_

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'T=

.  ??1 10 h = /,522 1C

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/oC2% Io '

l l 1

l N47 'L 4 S7$ $/2.Adf SO T ku ((one

=

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t n o I.475 &(v'*

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SHEET NO.

7!l9 JOB NO r?m W JOB + t/ A BY Nb DATE 1 %

CALC. NO, SUBJECT O t'> -' er Farryn '7,e c_. CHK'D DM DATE'7 ~ G ~ N 3CG SH-Oodd Ewn/Anr - / scr $vce.coun l'$A s e A - -

s2 T le.- CT'~ E I- l ] r- 5, + x), {

N.) ] ~

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et .

L J t k I.12 + C,1.s m S, ' + t A C,= (2. 400 %tv = 490 = I. lc,9 r

~400 / o$ 7 C C m

-o,I- I 7m l . I4 7 + 3 2h4 .

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m ~

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T z 0,9 T = >

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-.LAre./b~

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CF i O ,4 2f Ec J2./b s .

= 0 .6 O. AS*) Z 9 10 = l 7, O 76 ;w l

s23 /l .12f' Gancoa V C. =

/?D70/6*F?* 30

g

.- EQE INTERNATIONAL l aT-. >

l-SHEET NO. I7!N JOB NO. c ")~"-0C JOB " VA BY b DATE 7 ~~I' 4 I CALC. NO. . SUBJECT - &< ~ d 7eA / .' ' e " '" b'< CHK'D h DATE 1-i t -% ;

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i S

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= o . t.o!-,. z > 'io

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SHEET NO. I'l!I9 JOB NO. "2 2 07 0f JOB "T '/ A BY db DATE 1 -1 A4 r CALC. NO. SUBJECT Cem e' 7ee /b*'7 "'L CHK'D D h DATE 1 -t L'S k RG -SH-&d 4

L ,, c- d 6

• 0,4 EEr' L */ /6 - II 210 , 3 9.'

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EQE MERNATION4

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SHEET NO. A'/ft/

JOB No. S-n, , ,, JO,c ~r V A BY hl) DATE -r 7 - N CALC, NO. SUBJECT O + ' " ' / d' ' h= 7 bel CHK'D M DATE 'l- W SM G// cc44 k,7;!cp'tr: '

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l i

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m . s n . c o s a' f A2/it i

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& 603 7 " 24 El 799 9 \

El 7960 a 23 El 79/5 v 22 El 766.0 c= "

2/

o y El 770.5 c: "

/9 .

El 769.0 a: '#

o /7 '

El 759.5 c:

EQU/RWENT v is NATCH El 750./ c:

v /5 i

e /4 El 740.5 c: 1  !

~ 13 El 730.3 e:

~ tp i v ll El 72/5 ci ea v

9  :

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PERSON //EL "

8 LOCKS "

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\

4 El 69/./ d " 3 E/ 679.70 o

2 i

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,,v, '

l Figure 3.7.2-39 Steel Containment Vessel, Lumped Mass Model for Dynamic Analysis i

s k

{ ledoM tnemelE etiniF i

lesseV tnemniatnoC leetS 11-2 8,3 erugiF SRETNEC *4 NO

// w ,,., ,,//<

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,////, ,

l\'

LP"j/ f 87976 /

9 0/h lA ~ l 'e U96 lE SI ~ 4/

LP~kI = 5./07 lE 02 9/ LP"f,/

" = 53/7 lE 42~;32 LP "/ ;c 5./27 lE 03~i92 LP"! c 3037 lE 53-h 43 LP"f  := 5047 lE l 24~-\ /4 LP ~f

c

/057 lE od~;,74 LP-l ;c 5957 lE 65 55 tp -f :e O.967 lE 26~ l./6 LP 'j  := 5877 lE

/7~ 07 LP ",y 37 -

g 08./97 87 lE lE 47 ~, 67 77 / 5 0557 /E od 91 /

id 28 "

Lp -1 8997 lE 5d 4d V 7306 /E 76 6d 5906 lE hs 35/6 /E es SO LP'}

19 tp-f, 29 mLf P'E i

!t ltlsdF ddoc .dg . sec

Gcs .se- oedd f Af /st SQN-6

[

TABLE 3.7.2-13 STEEL CONTAINMENT VESSEL ELEMENT PROPERTIES l

E = 4,176,000 K/ft'; G = 1,670,400 K/Ft' i East-West Motion North-South Motion l Torsion Moment Moment i Element Length, Area, Constant, of Inertia, Shear of Inertia, Shear j No. Ft Ft' Ft' Ft* Factor Ft' Factor )

J 1 3.19. 44.40 1468 x 10' 734 x 10' 2.00 734 x 10' 2.00 i l

2 3.20 44.40 1468 x 10' 734 x 10' 2.00 734 x 10' 2.00 l I 3 5.71 43.75 1447.x 10' 734 x 10' 2.00 723 x 10' 2.00 4.00 3 4 39.81 1347 x 10' 674 x 10 8 2.00 674 x 10 8 2.00. '

l 5 5.62 38.70 1311 x 10' 656 x 10' 2.00 656 x 10' 2.00 t 6 4.08 39.64 1311 x 10' 656 x 10' 2.00 656 x 10 2 2.00 -

7 4.86 37.45 1239 x 10' 619 x 10' 2.00 619 x 10' 2.00 I 8 4.85 37.45 1239 x 10 2 619 x 10' 2.00 619 x 10' 2.00 i )

9 6.21 36.43 1202 x 10' 601 x 10' 2.00 601 x 10' 2.00 ig 1 10 3.50 39.43 1302 x 10' 651 x 10' 2.00 651 x 10' 2.00 I 11 5.50 29.36 970 x 10' 535 x 10' 2.00 435 x 10' 2.00  !

' M JS 12 4.00 27.96 970 x 10' 535 x 10' 2.00 435 x 10' 13 6.00 24.66 918 x 10' 514 x 10 2 2.00 405 x 10' 2.00 2.00 l l 14 3.50 25.31 918 x 10' 514 x 10' 2.00 405 x 10' 2.00 n N .15 6.50 25.09 867 x 10' 492 x 10' {

2.00 375 x 10' 2.00 l 16 9.00 24.02 794 x 10' 423 x 10' 2.00 371 x 10 8 2.00 i 17 6.33 24.02 791 x 10' 423 x-10' 2.00 371 x 10' 2.00 l 18 6.33 24.02 794 x 10' 423 x 10' 2.00 371 x 10' 2.00 l 19 6.34 24.02 794 x 10 8 423 x 10' 2.00 371 x 10 2 2.00 on 020 3.50 24.02 794 x 10' 423 x 10' 2.00 371 x 10' 2.00 21 6.00 18.06 597 x 10' 299 x 10' 2.00 299 x 10' 2.00 22 3.50 18.06 597 x 10' 299 x 10' 2.00 299 x 10' 2.00 23 4.52 16.17 535 x 10' 267 x 10: 2.00 267 x 10 8 2.00 24 10.38 15.93 512 x 10' 256 x 10' 2.00 256 x 10' 2.00 Urn 25 8.50 15.38 461 x 10' 231 x 10' 2.00 231 x 10' 2.00 26 10.25 13.08 321 x 10' 161 x 10' 2.00 161 x 10' 2.00 27 6.05 11.64 226 x 10 2 113 x 10 8 2.00 113 x 10' 2 00 Euta 28 9.80 9.35 117 x 10' 59 x 10' 2.00 59 x 10' 2.00 29 7.50 5.37 18 x 10' 9 x 10' 2.00 9 x 10' 2.00 l

0745F/COC4

Add.sM.codd P A5lal SQN-6 TABLE 3.7.2-14 STEEL CONTAINHENT VESSEL MASS POINT PROPERTIES Mass Total Weight Moment Eccentricity, Ft Point Weight, of Inertia North South East West No. Kips K-Ft' Motion Motion 69.51 2.30 x 10* 0.000 0.000 2 103.11 3.41 x 10 5 0.000 0.000 3 122.92 4.09 x 10 5 1.866 -3.510 4 149.68 5.02 x 10 5 3.063 -5.766 5 167.96 5.60 x 10 5 1.364 -2.569 6 125.28 4.14 x 10 5 0.000 0.000 7 104.66 3.46 x 10 5 0.000 0.000 8 140.99 4.66 x 10 5 0.000 0.000 9 160.48 5.30 x 10 5

-0.035 -0.324 10 132.02 4.36 x 10 5 0.078 0.759 l 11 118.63 3.97 x 10 5 0.703 8.770 12 166.93 5.72 x 10 5 0.666 14.909 l'6 13 179.67 6.16 x 10 5 0.850 11.455 l 14 147.60 5.01 x 10 5 2.173 9.381 1 6.37 x 10 5 15 190.89 1.168 5.712 16 199.49 6.59 x 10 5 0.657 0.542 17 151.76 5.02 x 10 5 0.715 0.414 18 131.23- '4.34 x 10 5 0.825 0.480

19. 104.76 3.46 x 10 5 0.821

[

0.617 20- 83.20 2.75 x 10 5 0.320 -0.208 'l-21 85.76 2.84 x 10 5

-0.095 -0.931 22 84.54 2.80 x 10 5 -0.085 -0.799 23 109.81 3.59 x 10 5 -0.038 -0.344 24 121.74 3.89 x 10 5 0.000 0.000 25 91.53 2.65 x 10 5 0.000 0.000 26 60.98 1.50 x-10 5 0.000 0.000 27 61.61 1.34 x 10 5 0.000 0.000 28 53.00 1.00 x 10' O.000 0.000 29 21 20 0.41 x 10' O.000 0.000 e

0745F/COC4

, scd.SH-oodd  ? A G/'t l SQN-6 TABLE 3.7.2-15 l

STEEL CONTAINMENT VESSEL PERIOOS i FOR NATURAL MODES OF VIBRATIONS North-South Motion Mode No. Frequency Participation (Hz) Factor i

1 9.346 1.717 2 16.129 -0.0001 3 23.810 -1.098 (p East-West Motion Mode No. Frequency Participation (Hz) Factor 1 9.259 1.731 2 16.129 0.004 3 23.810 -1.108 Vertical Motion Mode No. Frequency Participation (Hz) Factor 1 24.634 1.489 1

1 1

I 1

1 N.

0745F/COC4

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Figure 3.8.1-1 Reactor Building Elevation

' ' W sM- cedd PAa/t!

TENNESSEE VALLEY AUTHORITY 89/08/10.

SEQUOYAH NUCLEAR PLANT INTERIOR CONCRETE STRUCTURE EAST - WEST DIRECTION GROUND ACCELERATION = 0.18 G SPECTRAL EARTHOUAKE RESPONSE 5.0 PERCENT DAMPING HORIZONTAL DISPLACEMENT E-a 1

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{TACHMENT4 7 c/,,g. .fo TV/ 106'37 id[ CALCUbECiNO f(V/fW1"r// //o c./ Pag 31/, y TlTLE .%uo;e H'oog' ttda Sames,$ss m ,, Q ),,.

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BRANCH / PROJECT IDENTIFIERS Each time these calculations are issued, pIeparers must ensure that the original (RO) RIMS accession number is filled in.

,, g , , Rev (for RIM S use) RIMS accession number APPLICABLE DESIGN DOCUMENT (S)

R WD DK0800 nan R1 dA R2 g SAR SECTION(S) UNID SYSTEM (S)

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Yes ]No 5 Abstract yosA-- # eYl* A. ch .40 b 444.u 4 k n uL meryin i/cw o p f f , & .sm o .D e. Can da, - &% % .

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AS//3 Oh Record *G- M- 0047 hns.' by M B38 95011:0 804 Westinghouse Energy Systems Ba 355 Pittsburgh Pennsylvania 15230 0355 Electric Carparation Tek emccs -

TVA-95-002

1. MI67 b538 Th729 Sch Ref: MSE-REE-0785

2. 7v4W-/47. (538 44o8t3 Sbt) TASK N94-027

3. January 3, 1995 tvf444-IQ (339pogz4 po3)

Ll. TVA-%- 147 (m8 qqioot go _\

(on h d NO-Mr. Mark Burzynski, Manager 9 [QdP. 8(o'5C$$

Department'of Nuclear Engineering.

Tennessee Valley Authority k NH'O17 P. O. Box 2000 Soddy Daisy, TN 37379 0CC. 3 c b emcb(3eme -

TENNESSEE VALLEY AUTHOkITY SEQUOYAH NUCLEAR PIANT

[h Ice Condenser Beyond Desian Basis Evaluation-

Dear Mr. Burzynski:

This is a revised transmittal of our previous package on letter TVA-94-142, dated August 18, 1994.

Responses to EQE & Associates, Inc. Questions and Request of additional Information:

A. Top Deck Structure

1. The Dead Load (DL) Stress for the Radial Beam is 7.3 ksi.

l 2. The-DL Loads are; l

F, = 0 M, = 0 .

F y =0 My = -855.8 k-in F, = 10. 8 kips M, = +/- 0.1 k-in

3. The load coordinate system is; x direction - radial or perpendicular to the crane wall face.

y direction - tangential or parallel to the crane wall face.

z direction - vertical.

A copy of EDCE Sketch 72-633-A27, which also defines this coordinate system, is attached for additional clarity.

4. The cross section of the Radial Beam is that of a W10X77 Beam. EDCE sketch 72-633B-Sk2 is attached to further define the beam geometry.

5. The DBE Response Spectra, North-South Horizontal Direction 2% Damping, is defined on attached EDCE Drawing __

72-633B-A23.

6. The peak SSE acceleration occurs at the 8.87 Hz mode.

Other modes of the Structure at 5.99 Hz, 16.5 Hz, 20.4 Putnses in Performt.nce su.ma m

$cc, .sH-co47 A&fid Hz, 20.7 Hz, 28.0 Hz, 31.4 Hz, and 32.6 Hz. The modes in each excitation direction are combined by using either the " root sum square" approach for modes whose frequencies differ by more than 10%, or by the " absolute sum" for those modes whose frequencies differ by 10% or less.

7. The worst vertical DBE acceleration is 0.58g and the worst horizontal DBE acceleration is 4.55g, at 8.87 Hz for the N-S Horizontal Response Spectra.

8. The Top Deck Structure, in our judgement, is one of the critical support structures of the Ice Condenser in regards to having limited margin beyond the original design and failure of this structure could limit the Ice Condenser from performing its intended safety function.

B. Ice Baskets

1. The following failure load combinations were established from the qualification testing program; Horizontal (Lbs) Vertical (Lbs) a) 1176 5339 b) 1721 3940 c) 1048 5100 C. Crane Wall Cradles

1. The seismic inputs into the Cradle are the reactions through the lattice frame attachment brackets, which is the result of the reactions of the ice baskets (including impacting loads) onto the lattice frames, and the seismic excitation of the lattice frames (each lattice frame eights 1200 Lbs).

2. The Crane Wall Cradles were tested to failure under a combined radial SSE load of 36 kips plus a radial DBA load of 16.5 kips, which equates to design loads of 24,000 Lbs (SSF) and 11,000 Lbs (DBA), respectively. The DBA loads were applied uniformly to the horizontal support beam, while the SSE loads were applied to the cradle ear at the end on the horizontal support beam.

D. Lower Support Structure

1. The Radial Beams are a weld fabricated rectangular section made from a 10X25 Channel Section and a 1/4" '

plate.,

2. The Radial Beams are attached to the Inner and outer l Circumferential beams with a 5/16" and 3/8" fillet veld all around.

3. We have not been able to identify the specific weld wire (E60 or E70) used to attach the radial beam to the Circumferential beams. This information may be on the

. Pittsburgh Bridge & Iron Co. Weld Procedures and/or Shop Fabrication Drawings which have been sent to mine storage. PB&I is no longer in business. The design allowable stress for the weld material is 21 ksi.

4. From the Stress Reports we have been able to retrieve we J

SM-SM - 004'7 d 7,li3 l

1 I

have not been able to identify, oc separate out, the Dead

Weight (DL) Loads on the individual Radial Beam Members.

l We bave~been able to identify the design dead weights of all the structures supported by the LSS, which are; a) Ice Basket weight - 2000 Lbs each basket b) Wall Panel weight -

4000 Lbs per Lattice frame (1/3 bay) section c) Lattice Frame weight -

9600 Lbs per 8 Lattice Frame bay d) Intermediate Deck - 2200 Lbs per LSS bay e) Lattice Frame Columns - 989 Lbs each column

5. The lower Support Structure, in our judgement, is one of the critical support structures of the Ice Condenser in Regards to having limited margin beyond the original design, and failure of this structure could limit the Ice Condenser from performing its intended safety function.

6. The North-South DBE Horizontal Response Spectra for Spectral Mass Point 7 caused the highest loadings on the LSS.

7. The Radial SSE plus Vertical SSE load combination produces the worst case design loadings.

E. Lattice Frame Crane Wall Columns

1. The cross-section of the column is a solid bar 2.0" X 4.88".

F. Lattice Frames

1. The Lattice Frames, in our judgement, are one of the critical support structures of the Ice Condenser in regards to having limited margin beyond the original design, and failure of this structure could limit the Ice Condenser from performing its intended safety function,

2. The cross-section of the radial lattice member no. 71 is 1/2" x 7 1/2".

Attachments: EDCE Dwg 72-633-A27 72-633B-SK2 72-633B-A23 Tables (3 sheets) - DB Stress Information very truly yours, WESTINGHOUSE ELECTRIC CORPORATION

'r' '

D. W. Sa k Sequoyah Project Manger TVA Projects cc: D. Lafever

SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION PEge 1 of 3 Rev. 3, 12/22/94 .

Top Deck Structure Ice Brr.kets Crane Wall Cradles Critical Component Radial Beam Perforated Basket Cradle Arm / Side ,

Cylinder Sheet Metal Plate / Insulator Bushing / Bolt Interface 8 Critical Location Crane Wall Connection 6 Foot Elevation above 45 Foot Elevation above

. LSS LSS raulted Loads SSE DL+DBA D+DBA+SSE* SSE*

r, = +/- 1.7 kips r, = +.3/ .2 kips 921 Lbs Horizontal 11200 Lbs Radial 2409 Lbs Vertical r,=+/-7.0 kips r,=+.3/ .2 kips DBA*

7837 Lbs Radial r,= + /-11. 3 k i p s r,=33.3 kips ,

M,= 0.0 k-in M, = 0.0 k-in M, = +/- 133.3 k-in M, = -2679.5 k-in M, = +/- 623.6 k-in M, = +4.2/-82.7 k-in Total Paulted Stress 41.2 kai (DL+DBA+SSE) N/A Qualification by N/A Qualification by test test Normal Operating Stress or Load 21.9 kai (DL+LL) 0 Lbs Morizontal' N/A 1571 Lbs Vertical

• SSE Stress or Load 26.6 kai (DL+SSE) 385 Lbs Horizontal 11200 Lbs Radial 864 Lbs Vertical *2 Modes 3.52 Hz-Horizontal >33 Hz >35 Hz Response Spectra or Spectral Acc. Interior Concrete Mass Point No.14, 0.55g vertical 0.55g Vertical Daeping .02, DBE E/W & N/S Material ASTM A441 ASTM A569 ASTM A441 Allowable stress or Loada 36 kai (DL+SSL) o,i m a,= 1.33 x 0.6r, 2246 Lbs Horizontal Ultimate Test Load 45 kai (DL+DBAv5SE) o.a w s.= 1.65 x 0.6r, 313a Lbs Vertical

• Combination to railure:

24000 Lbs Radial (SSE) 11000 Lbs Radial (DBA) r,= Min. Yield Strength * $ critical Location

• e critical Location

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SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION P&ge 3 af 3 m.v. 3, 12/22/ 4 i

Lattice Frames Lattice Frames Critical Component Moeber 71 Member 71 Critical Location 15 Foot El.(faulted) 57 Foot El. (seismic) ,

Faulted Loads DBA+SSE 15' El. DBA+SSE 57' El, I

12321 Lbs Vertical 3420 Lbs Vertical 14809 Lbs Tangential 13338 Lbs Tangential 6211-Lbs Radial 0 Lbs Radial or or 3420 Lbs vertical  !

12321 Lbs vertical 13824 Lbs Tangential 11896 Lbs tangential 11776 Lbs Radial 8148 lbs Radial e or or l 12321 Lbs vertical 3420 Lbs Vertical [

8259 Lbs Tangential 3748 Lbs Tangential  !

15381 Lbs Radial 13426 Ibs Radial Total Faulted Stress (max) 49330 psi * (@l5' ele.) Not Calculated 0 Lbs Horizontal 0 Lbs Horizontal  !

Normal Operating Stress or Load 1200 Lbs vertical _

1200 Lbs Vertical [

t SSE Stress or Load 15 Foot Elevation 57 Foot Elevation 660 Lbs vertical 660 Lbs vertical f 6550 Lbs Tangential 9590 Lbs Tangential 0 Ibs Radial 0 Lbs Radial i I

or or 660 Lbs Vertical 660 Lbs Vertical 5565 Lbs Tangential 8148 Lbs Tangential 5565 Lbs Radial 8148 Lbs Radial j or or 660 Lbs Vertical 660 Lbs vertical 0 Lbs Tangential O Lbs Tangential 9170 Lbs Radial 13426 Lbs Radial ,

l Modes >35 Hz >35 Hz Response Spectra or 0.55g vertical 0.55g vertical {

g Spectral Acc.

Material ASTM A441 ASTM A441 h!

Allowable stress or Loads 61875 psi 61875 psi  ;

cas t = 1.65 x 0.75F, casat,= 1.65 x 0.75F, o i

• @ critical Location k N

• • Lattice Frame Load testing and Qualification testing was not performed to failure.

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TVA:IPEEE Probabilistic Analysis-Sequoyah Interior Concrete Structure A Node 1014 CM Elev. 796.63 ', Translation in EW - (y) Dir. tm

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Legend: Notes:

3% Damped Spectrum Input Median CR0098 Shape 5% Damped Spectrum __. __ _ _ _

- (Ensemble at Rock Outcrop) 7% Damped Spectrum _________ Accelerations in g's 4 84th Percentile Spectrum i 8

TVA:.IPEEE Probabilistic Analysis-Sequoyah Interior Concrete Structure A Node 1014 CM.Elei. 796.63, Translation in EW (y) Dir. $;

e

_ _ _ _ = - _ . _ _ _ _ _ _ _ _ _ _ _ . _ -_-_____ ---__ - ______ __ - - ___. - ._ - _ .- _ . .- . . . . -

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( h5.' & B38 95011:0 804 Westinghouse Energy Systems Ba 355 Pinsburgh Pennsytvania 15230-0355 Electric Corporation Te(c<emcc5'.

TVA-95-002

1. MI62. d538 Th729 Sot) Ref: MSE-REE-0785

2. rvA-W-lu(B3s 4@8u set) TASK N94-027

3. -Tvl4 94-I4p ($$591(o826 po3) January 3, 1995

4. 7YA W- 27 dn8 Syroot goq} j (onkmch No.

Mr. Mark Burzynski, Manager 9 lMP- 8(o3o5B Department of Nuclear Engineering Tennessee Valley Authority y

P. O. Box 2000 Soddy Daisy, TN 37379 dCC 3 e b( emdo33 Hz >35 Hz Response Spectra or Spectral Acc. Interior Concrete Mass n'oint No.14, 0.55g vertical 0.55g Vertical Damping .02, DBE E/W & N/S Material ASTM A441 ASTM A569 ASTM A441 Allowable stress or Loads 36 kai (DL+SSE) coa = 1.33 x 0.6r, 2246 Lbs Horizontal Ultimate Test Load 45 kai (DL+DBA+SSE) cones.== 1.65 x 0.6r, 3L38 Lbs Vertical

• Combination to railure:

24000 Lbs Radial (SSE) 11000 Lbs Radial (DSA) a O r,= Min. Yield Strength

• e critical Location e critical Location q D.

1 0

0 A

A k

s

Ptge 2 af 3 SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION Rev. 3, 12/22/94 Lattice Frame Crane Wall Crane Wall p nts .

Lower Support Structure (LSS)

Columns End Column Collar Stud (TVA Scope) critical Component Inner Radial Beam weld Connection to Middle 33' Elevation above LSS 12' Ele

• for SSE Loads ,

Critical Location Circumferential Beam 36" Ele

• for DBA and DBA+SSE 144958 Lb-in* DBA+SSE raulted Loads (D+DBA+SSE) D + DBA* (seE SSE Loads below) 11.)$ kips (Radial) 14.16 kips (Tangential) r,= 6.3 kips r,= -1.1 kips r,= -19.4 kips M,= -1.3 k-in t M,= -653. 9. 4 k-in M,= 20.0 k-in 52165 pai* TVA Scope Total raulted' Stress 34.3 kai*

4300 pai (estimated) TVA Scope ,

Norma'l Operating Stress or Lead 14.8 kai 1 77000 Lb-in (estimated) 14 kips Radial SSE Stress or Load SSE Loads

• 10 kips Tangential  ;

r,= -5.2 kips r,= 0.4 kips r,= 14.6 kips M,= 9.0 k-in M,= -298.4 k-in M,= 19. 2 k-in

>35 Hz >35 Hz Modes 14.1 Hz N/A (0 0.55g Vertical 0.55g Vertical Response hectra/ Spectral Acc. 0.92g Radial 0.88g Tangential i ,

ASTM A434 (/)

ASTM A588 ASTM A588 Material 2 55687 psi 50 ksL min yield ,

Allowable Stress / Load 36 kai TVA Scope g c as .= 1.33 x 0.6r, o g i%i,= 1.65 x 0.15r, D

• e critical Location

• Elevation above LSS

• O critical Location &

x h

w

rog. 3 of 3 SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION Rev. 3, 12/32/94 P Lattice Frames Lattice Frames 'I Critical Component Member 71 Member 71  ;

Critical Location 15 Foot El. (faulted) 57 Foot El. (seismic) , l Faulted Loads DBA+SSE 15' El. DBA+SSE 57' El.

l 12321 Lbs Vertical 3420 Lbs Vertical 14809 Lbs Tangential 13338 Lbs Tangential 6211 Lbs Radial 0 Lbs Radial l

or or 12321 Lbs vertical 3420 Lbs Vertical 13824 Lbs Tangential 11896 Lbs tangential 11776 Lbs Radial 8148 lbs Radial or or 12321 Lbs Vertical 3420 Lbs Vertical 8259 Lbs Tangential 3748 Lbs Tangential 15381 Lbs Radial 13426 Lbs Radial Not Calculated Total Faulted Stress (max) 49330 psi * (@l5' ele. )

0 Lbs Horizontal 0 Lbs Horizontal Normal Operating Stress or 1200 Lbs Vertical Load 1200 Lbs vertical 15 Foot Elevation 57 Foot Elevation SSE Stress or Load 660 Lbs Vertical 660 Lbs Vertical 6550 Lbs Tangential 9590 Lbs Tangential 0 Lbs Radial 0 Lbs Radial or or 660 Lbs Vertical 660 Lbs vertical 5565 Lbs Tangential 8148 Lbs Tangential 5565 Lbs Radial 8148 Lbs Radial or or 660 Lbs vertical 660 Lbs Vertical 0 Lbs Tangential 0 Lbs Tangential 9170 Lbs Radial 13426 Lbs Radial

>35 Hz >35 Hz Modes 0.55g vertical 0.55g vertical g Response Spectra or Spectral Acc. Q .

ASTM A441 g Material ASTM A441 61875 psi Allowable stress or Loads 61875 psi i,= 1.65 x 0.75F, au t.,,.i.= 1.65 x 0.75F, a ut

• @ critical Location

• • Lattice Frame Load testing and Qualification testing was not performed to failure. y

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g g ,, j gmg,fg g4 PLANT / UNIT TITLE na. e..su w w/g' 46 g sau/ i n PREPARING ORGANIZATION KEY NOUNS (Consult CCRIS LIST)

DEleoE Sei.v>a., 27L"eEi 6 dwy /, Euui"w 9 c BRANCH / PROJECT IDENTIFIERS Each time these calculations are issued, prepders must ensure that the oriDinal(RO) RIMS accession number is filled in, RIMS accession number gg, y og Rev (for RIM'S use) hWW Dx0809 osa APPLICABLE DESIGN DOCUMENT (S) R1 y* R2 N

W SAR SECTION(S) r)A UNID SYSTEM (S) 4 A- (, / R3 -

Revision 0 R1 R2 R3 Safety-related? Yes No EGN No. (or indicate Not Applicable)

AJA Statementof Problem f 4Lb b rr%G YMf }/'Ct/f /ttbet fy- tb &<* - Checked Y 1 * #

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TV Us es deleted by this revision jf(j[j C M8[~ [ w c. Ust all pages changed r=4**1 by this revision These calculations contain unverified assumption (s) Calculation contains special requirements or that must be verified later. Yes No limiting conditions. Yes No g ) Abstract a.hrAf E M@ M MEC--

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SCG -St-f- 00 49 o l

                       ' Calculation No.                         Revision                                                    !

I Method of design verification (independent review) used (check method used):

1. Design Review

2. Altemate Calculation

3. Qualification Test 1

Comments:

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         %4 L':9 C'S                                                                        SHEET NO. 4/'/a d

JOB NO. 62 2o7 JOB Y A- Shhl Isn'M/c MA4d/M BY O8/3 DATE I f CALC. NO. SUBJECT bI / Anc9' IeAsn 5>PAoM N rYECHK'D SL DATE '/13/G6 ! W-GPl-0049 dLE oF /75+f 7't. (JE; CCEIP770 fA hg Covee GHerr ( l l-s ST of SM/s10H S 2 bN G57'csMd w-HY {&Ie u) foR. M' 3 [A d L E OF 0/f 73rH 7C $ R w a.cnewe s~  ; Sumne v C A s s me nuis c,  ; Irm e Co vare a 4 D e-c u an r1oN t 7 - /0 N C-L U C/ O N l0 1 i Yy !!! b (/f h e.sh j ' i ( l l j l

i  ; 1 l EGE see mmnarieuxt S I.

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         ;os No. sz2,7            aos      T vA      s a s/                           By _dat)EET     No. /93-DATE f CALC. No.                SUBJECT LoueT        Iu PFo PT f72 0 c.:7 ,         CHK'D SL     DATE l/I N5 l

sc4 -SH- oc49 I?_ r P e 1 a v c r c , 7 v v w f egde>%N [ce 0MOwHret h65 t c M { ro x, Twoenmsu d w .swcm (we$Ast.Snawossefer I rlu z+ >'NsIc T ' (T t/A.) Paa n ase cs n H ou,envis x ejig 74 j 7K4-99-/#3i A 18 A4 0823 80/ At G- VAvI 2, s;eavova n.- wonesa. su,u,&cii % ,cc TupoiawA 77o u (2.ani .2 X ,1%aa l'rVA) 70 I . Ku er-0 ( WG~) ,1 amson.4c- Ce $xuwies770M , 49/ rl1 l 3, "4 Meruosotoc, y ioeA --s s a - ~ r or - ,x /o c c - <e Pococr % nT S'ersnic /%e,ia' 12av 1. A u c, i99i, C E PRI. EEPoAT 6J P - Go4-l- SL) 9 ED E G e e, Alo Cazo 7,0/ -C - co 4 )2mto " kemcre \([*] reew rn - k .oer o a.e levens-

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\ h l l' W l k < f f9pce] VJr 0)f l , N3/52 ($38 9 9 0 71.9 801), \ 7V,4 9 9-/44 618 9+ 06.26 603 l w x JC G- G/%r l v MMAr-.y l

                            ~i~Hc Coarzou.inc. R Ls + Hot crezecs is                         zs, I    l- u' to t'r7f      A      De<sia a Acto es . S.rzess              or      z&    k v' F S.        =    l . 03                                                                     l GM A =        F 5. ( ggr ')     =

l o 2 ( 0. y ) = 0.?>l g. O

M l y-D. EQEINTERNATONAL

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SHEET NO. d' o JOB NO. 67207 JOB TYA SM N BY Nd/>) DATE I/4 I CALC. NO. SUBJECT [ n t > r7t $o P t'oe T Sm v LT' CHK'D SL DATE /f3 ' 6t% Std 00f9 j l C $ v M r2 7~/ 0 H C D Csa r77 o c <xe, [ r>r.e e c /s / % vr.s ex- [pz s .c s  ! da r#erz. 77/e# 6t.)sza STP_e C5 ' herr _o d w r#c/ods ]

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( on< v cua m.7 He n ra.%6;,eonce vm

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• flor Co e c en an ,,v e Mav na .

Tk in C 0c / e2 vn h ( Duee Eno,s u &n ) j 1 l

4 Mm - EQE WTERNATIONAL I .

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SHEET NO. 7/'o JOB NO, S'2247 JOB WA - GOh/ $ /r w/c M 4f2.a/A/ BY AO/ DATE // /7 1 CALC. NO. SUBJECT 5s A r luT [mW (OPPoe7 (7PucTCHK'D SL DATE I /I3 1E us -sM- ce47 Lo wcz [urroer 6mucwes fl~ma /2anuu. 16x l OL + O gA F SE 34.3 hcE Caus CL- 19.8 hd 1 CL+ O R A + c Ct. G< G.1 L F .- -I.I k r r -13.A b , hf,_ : - la Z . b-w

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SsG Fx = - S' z k F, 7 o,4 k  ! Q_ r / 4 . 6. k .

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                                                                                       =         13 2     hA dL + dd A                                 $ .:-        4,3 - [- 9,2 ) r                     //,8     l<

F, = -l,l - 0 ,4 = - lac h

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pi : g ws,9 - (-zy.4) - - 2ss, c k .-m l JMz: 20.0 - 19 2 = 62 k-m s i

N.- EQE INTERNATIONAL a K4 7

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SHEET NO, b/D JOB NO, 622o7 JOB T\/A BON BY - Odb DATE I( /7, ' CALC, NO. SUBJECT b o w n- So Pi*o er $ TE ue r CHK'D SL DATE - l/G/15

                                      $4;.SH w49 2 A c / A L-         L'A M              E   10x25          +     0. 2 S'     fcr.

[ I ' 30.7 Y kr. 5- I lt O,tL~ /0 - 20 2 i 90 .7 + ZD .E = lll.5 N

                                                                                                          . 2 C=        $                 $'

x 2E,I w T h I .f LJ.- M.2 <s =- Ma n.L . rx, = 3so LL fana T ( 6 c H C f1A & C S C = M,. . tsz. ? - zv. 3 l<ei E% i z .1 [T- 3.1 $l > Pu- A- c.er 10 - z.c l' c

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                                         ~

G 4,72312 k- , I d M' X = 0.6d w I .2 i A,: 9.zz a 2. c : 9.2: - i 6

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        ,g 1. EQE INTERNATIONAL gm Y
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SHEET NO. Y/ o l JOB NO. 522o7 JOB TVA 5 O Al By d4 d DATE 11/7 CALC. NO. SUBJECT /m via' S up Ao e 7- frE u eT- CHK'D SI- DATE I/I3 GF Sc6-s+ ev41 l lu e 3.4+ 2.151 'O.0/Z &(< 304

                                                                                              /I, c 7      OY

( c= { 2 . E 7s - ), IG L } = /, 7/ 3 L 57 4.93 - I 5, 75 k. c, ' l CT~ = 20.O + Q = z,ep + 0.c4 -

4. >I 943.

N A65 C (210!!A t CLer

29. 3 F z.s ?. s .3 L. 2 k-l [vc. 11 1 ) '

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  ,        Assuna                  C<sse         <wuee.                                                           en   asc.s 5rn.<    1.
                   .    -       CSE          C772tCC U =           2 *)2. 4     +        19.2       L S,2               w      }&. 7     l.c           A t2 S.

Z E.3 6,93 *) . El

/3,7 " CR G

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Yonn CfY = 0.22 1 e I A, 7 He { deva o bo m e.  % (v - o,9t ' Ye7e $g ( E) - C <5 e c Asssua s x c., rnric At a rissa iacs (re 7et,>e-S ?o 0 MP. / 'l . 7 lI4 [ l$ cYn Yn lucrioriG ' IIc c.tE. .

                        $2 $7    0 Sel'       k                 ~ C. b           I; YeH$o                \/ A T : hr       FRS E- kl       (L z 0. 4 , f I c.              Rso AL T*               :

B d . 0.74 f- _1 9. 2 ,o.S . 5.2 .oA2 - zo.2 LJ Zz.z o,rs c .93 o.zz 9ar o,x

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i l . l M - EQE 24TERNATONAL i I 4 ) (%? .h

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l i SHEET NO.10,//o JOB NO. 52207 JOB T1/A 4 t A/ By V/d6) DATE / / M M i CALC. NO. SUBJECT L/A iarr Co 7p o rr r $77to e_ r CHK'D SL DATE 8/83 ' 5)'

                          && -5//-0041 20,Z L / 4,2                                                  =             35.I        Ncl          C     36              NO h1. S .                           >               ).O 3                                    OK 0 Mci vsi on

, ns is G es n o riva in sur //oa. is Fia-e o is in rw <2. , c, i o  % as . .Tr  ; ! ivn y Mor- re csss evs vi ve if ras 0 ye rno s. ! YYo (2 eC~r k R. I t PprA. /C /N THE Z0 Ab 2nNc,E. [ on et i4.7 fls % an Giyen W /lo d / g. e-crion i l j l l l

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SHEET NO. A/ [ JOB NO. 52207 JOB T / A - $ 8 M I W 5 m /c dden Marb4) BY k)M/ DATE I 4 CALC NO. SUBJECT 7td 6nbs42w &iner S ce<lJ/xe 'AA-- CHK'D SL DATE \ h3A15 Sca 5M-oo49 l i X ffen l Arfe i

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09/15/1994 09:11 CIVIL ENGR DSA 1R-SON 615 751 0247 P.10 G'A ~ '*.M- 00 <f 9 ,.,* d2 i .}

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AW QR Record S h ,a-w/#on.9 ( nS.' & B38 950110 804 Westinghouse Energy Systems $3lje enum m30.cass Electric Corporation TeScA emccS'. TVA-95-002

1. Nll62. d538 Th724 Sch s Ref: MSE-REE-0785

2. T/ A-W- /% (1538 4@823 56t) TASK N94-027

3. -N4.cni-19 cgw.7goggg, po3} January 3, 1995 Ll. 'DIA W- 147 (m8 qqioot goq) 0anb4Cb-h;o.

Mr. Mark Burzynski, Manager 9 ldh)P- 8(i3CS B Department of Nuclear Engineering # Tennessee Valley Authority k NH~O17 P. O. Box 2000 Soddy Daisy, TN 37379 dCC- 3 e foc emcb<semN. TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT [h Ice Condenser Beyond Desian Basis Evaluation l 1

Dear Mr. Burzynski:

This is a revised transmittal of our previous package on letter TVA-94-142, dated August 18, 1994. Responses to EQE & Associates, Inc. Questions and Request of additional Information: A. Top Deck Structure l

1. The Dead Load (DL) Stress for the Radial Beam is 7.3 ksi. l l

2. The DL Loads are; F, = 0 M, = 0 F y =0 My = -855.8 k-in F, = 10. 8 kips M, = +/- 0.1 k-in

3. The load coordinate system is; x direction - radial or perpendicular to the crane wall face.

y direction - tangential or parallel to the crane wall face. z direction - vertical. A copy of EDCE Sketch 72-633-A27, which also defines this coordinate system, is attached for additional clarity.

4. The cross section of the Radial Beam is that of a W10X77 Beam. EDCE sketch 72-633B-Sk2 is attached to further define the beam geometry.

5. The DBE Response Spectra, North-South Horizontal Direction 2% Damping, is defined on attached EDCE Drawing __

72-633B-A23.

6. The peak SSE acceleration occurs at the 8.87 Hz mode.

Other modes of the Structure at 5.99 Hz, 16.5 Hz, 20.4 Pg, nn in Pedorrv.nce a as n-a m

4 47[/

                                                              $c.4 -s H - oo4'9 Hz , 2 0. 7 Hz , 28. 0 Hz , 31. 4 Hz , and 3 2. 6 Hz. The modes in each excitation direction are combined by using either the    " root   sum    square" approach for modes whose frequencies differ by more than 10%, or by the " absolute sum" for those modes whose frequencies differ by 10% or less.

7. The worst vertical DBE acceleration is 0.58g and the worst horizontal DBE acceleration is 4.55g, at 8.87 Hz for the N-S Horizontal Response Spectra.

8. The Top Deck Structure, in our judgement, is one of the critical support structures of the Ice Condenser in regards to having limited margin beyond the original design and failure of this structure could limit the Ice Condenser from performing its intended safety function.

B. Ice Baskets

1. The following failure load combinations were established from the qualification testing program; Horizontal (Lbs) Vertical (Lbs) a) 1176 5339 b) 1721 3940 c) 1048 5100 C. Crane Wall Cradles

1. The seismic inputs into the Cradle are the reactions through the lattice frame attachment brackets, which is the result of the reactions of the ice baskets (including impacting loads) onto the lattice frames, and the seismic excitation of the lattice frames (each lattice frame

          ,  eights 1200 Lbs).

2. The Crane Wall Cradles were tested to failure under a combined radial SSE load of 36 kips plus a radial DBA load of 16.5 kips, which equates to design loads of 24,000 Lbs (SSE) and 11,000 Lbs (DBA) , respectively. The DBA loads were applied uniformly to the horizontal support beam, while the SSE loads were applied to the cradle ear at the end on the horizontal support beam.

D. Lower Support Structure

1. The Radial Beams are a weld fabricated rectangular section made from a 10X25 Channel Section and a 1/4" plate. .

2. The Radial Beams are attached to the Inner and outer Circumferential beams with a 5/16" and 3/8" fillet weld all around.

3. We have not been able to identify the specific weld wire.

(E60 or E70) used to attach the radial beam to the This information may be on the Circumferential beams. Pittsburgh Bridge & Iron Co. Weld Procedures and/or Shop Fabrication Drawings which have been sent to mine storage. PB&I is no longer in business. The design allowable stress for the weld material is 21 ksi.

4. From the Stress Reports we have been able to retrieve we

 .                                                           Scc. -SH- ood? A8lll have not been able to identify, or separate out, the Dead Weight (DL) Loads on the individual Radial Beam Members.

We have'been able to identify the design dead weights of all the structures supported by the LSS, which are; a) Ice Basket weight - 2000 Lbs each basket b) Wall Panel weight - 4000 Lbs per Lattice frame (1/3 bay) section c) Lattice Frame weight - 9600 Lbs per 8 Lattice Frame bay d) Intermediate Deck - 2200 Lbs per LSS bay e) Lattice Frame Columns - 989 Lbs each column

5. The lower Support Structure, in our judgement, is one of the critical support structures of the Ice condenser in Regards to having limited margin beyond the original design, and failure of this structure could limit the Ice Condenser from performing its intended safety function.

6. The North-South DBE Horizontal Response Spectra for Spectral Mass Point 7 caused the highest loadings on the LSS.

7. The Radial SSE plus Vertical SSE load combination produces the worst case design loadings.

E. Lattice Frame Crane Wall Columns

1. The cross-section of the column is a solid bar 2.0" X 4.88".

F. Lattice Frames

1. The Lattice Frames, in our judgement, are one of the critical support structures of the Ice Condenser in regards to having limited margin beyond the original design, and failure of this structure could. limit the Ice Condenser from performing its intended safety function.

2. The cross-section of the radial lattice member no. 71 is 1/2" x 7 1/2".

Attachments: EDCE Dwg 72-633-A27 72-633B-SK2 72-633B-A23 Tables (3 sheets) - DB Stress Information very truly yours, WESTINGHOUSE ELECTRIC CORPORATION D. W. SaMk [ ' sequoyah Project Manger TVA Projects cc: D. Lafever m

SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION Page 1 af 3 Rev. 3, 12/22/94 Top Deck Structure Ice Baskets Crane Wall Cradles Critical component Radial Beam Perforated Basket- Cradle Arm / Side Cylinder Sheet Metal Plate / Insulator Bushing / Bolt Interface 8 Critical Location Crane Wall Connection 6 Foot Elevation above 45 Foot Elevation above LSS LSS SSE DL+DBA D+DBA+SSE* SSE* raulted Loads 921 Lbs Horizontal 11200 Lbs Radial F,= +/- 1.7 kips r, = +.3/ .2 kips 2409 Lbs Vertical r,=+/-7.0 kips r,=+.3/ .2 kips DBA* TEST Lbs Radial r,=+/-11.3 kips r,=33.3 kips , M,= 0.0 k-in M, = 0. 0 k-in M, = +/- 733.3 k-in M, = -2619.5 k-in M,= +/- 623.6 k-in M, = +4.2/~82.7 k-in Total raulted Stress 41.2 kai (DL+DBA+SSE) N/A Qualification by N/A Qualification by test test Normal Operating Stress or Load 21.9 kai (DL+LL) O Lbs Horizontal' N/A 1511 Lbs Vertical

• SSE Stress or Load 26.6 kai (DL+SSE) 385 Lbs Horizontal 11200 Lbs Radial 864 Lbs Vertical *2 Modes 3.52 Hz-Horizontal >33 Hz >35 Hz Response Spectra or Spectral Acc. Interior concrete Mass Point No.14, 0.55g Vertical 0.55g vertical Damping .02, DBE E/W s N/S Material ASTM A441 ASTM A569 ASTM A441 Allowable stress or Loads 36 kai (DL+SSE) o.aaes.= 1. 33 x 0. 6F, 2246 Lbs Horizontal Ultimate Test Load ,

45 kai (DL4DBA+SSE) 0.3g 3,= 1.65 x 0.6r, 3138 Lbs Vertical

• N hination to railure:

24000 Lbs Radial (SSE) 11000 Lbs Radial (DBA) r,= Min. Yield Strength

• O critical Location
• e critical Location I

h 1 O A s

POge 2 of 3 SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMATION Rev. 3, 12/22/94 Lower Support Structure (LSS) Lattice Frame Crane Wall Crane Wall Embedmonte

• Columins Inner Radial Beam End Column Collar Stud (TVA Scope)

Critical S nent 33' Elevation above LSS 12' Ele

• for SSE Loads - .

Critical Location Weld Connection to Middle Circumferential Beam 36' Ele

• for DBA and DBA+SSE D + DBA* (see SSE Loads below) 144958 Lb-in* DBA+SSE raulted Loads (D+DBA+SSE) 11.15 kips (Radial) r,= 6.3 kips 14.16 kips (Tangential) r,= -1.1 kips r,= -19.4 kips M,= -1. 3 k-in M,= -653. 9. 4 k-in M,= 20.0 k-in 34.3 kai* 52765 pai* TVA Scope Total raulted Stress Normal Operating Stress or Load 14.8 kai 4300 poi (estimat ed) TVA Scope SSE Stress or Load SSE Loads
• 77000 Lb-in (estimated) 14 kips Radial 10 kips Tangential r,= -5.2 kips F,= 0.4 kips r,= 14.6 kips M,= 9.0 k-in M,= -2 9 8. 4 k-in M,= 19.2 k-in 14.7 Hz >35 Hz >35 Hz Modes Response Spectra / Spectral Acc. 0.55g vertical . 0.559 vertical N/A 0.92g Radial g 0.88g Tangential n D

ASTM A588 ASTM A588 ASTM A434 , Materini O 55687 psi 50 kai min yield Allowable Stress / Load 36 kai o, ..= 1.65 x 0.75r, TVA Scope d o,i nwi,= 1.33 x 0.6r, n

                                                                                      * @ critical Location

• Elevation above LSS D
• 9 critical Location 0-hs t

rage 3 at 3 SEQUOYAH ICE CONDENSER DESIGN BASIS STRESS INFORMkTION m.v. 3, 12/22/34 Lattice Frames Lattice Frames . Critical Component Member 71 Member 71 Critical Location 15 Foot El.(faulted) 57 Foot El. (seismic) , Faulted Loads DBA+SSE 15' El. DBA+SSE 57' El. 12321 Lbs vertical 3420 Lbs Vertical 14809 Lbs Tangential 13338 Lbs Tangential 6211 Lbs Radial 0 Lbs Radial or or 12321 Lbs vertical 3420 Lbs ver*4 .,1 13824 Lbs Tangential 11896 Lbs ta vential 11776 Lbs Radial 8148 lbs Radial . or or 12321 Lbs vertical 3420 Lbs vertical 8259 Lbs Tangential 3748 Lbs Tangential 15381 Lbs Radial 13426 Lbs Radial Total Faulted Stress (max) 49330 psi * (@l5' ele.) Not Calculated Normal Operating Stress or 0 Lbs Horizontal 0 Lbs Horizontal Load 1200 Lbs vertical 1200 Lbs vertical SSE Stress or Load 15 Foot Elevation 57 Foot Elevation 660 Lbs vertical 660 Lbs Vertical 6550 Lbs Tangential 9590 Lbs Tangential 0 Lbs Radial 0 Lbs Radial or or 660 Lbs Vertical 660 Lbs vertical 5565 Lbs Tangential 8148 Lbs Tangential 5565 Lbs Radial 8148 Lbs Radial or or 660 Lbs vertical 660 Lbs vertical 0 Lbs Tangential 0 Lbs Tangential 9170 Lbs Radial 13426 Lbs Radial (A Modes >35 Hz >35 Hz Response Spectra or 0.55g vertical 0.55g vertical h Spectral Acc. ASTM A441 ASTM A441 o Material s 61875 psi A Allowable stress or Loads 61875 psi

• a,u 1,= 1.65 x 0.75F, o,n 1,= 1.65 x 0.75F,

                                                              * @ critical Location b
                  **        Iattice Frame Load testing and Qualification testing was not performed to failure.                                                                                                                                                                  4 x

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TVA10697 db,T!CNO /g Page 1 PLANT / UNIT TITLE IPEEE Seismic Margin Evaluation Auxiliary Control Building SQN 1/2 PREPARING ORGANIZATION KEY NOUNS (Consult CCRIS LIST) NE/EQE Seismic, Margin, IPEEE Each time these calculations are issued, preparers must ensure that the original (RO) RIMS accession number BRANCH / PROJECT IDENTIFIERS is filled in. v U Rev (for RIM'S use) RIMS accession number g

                                                                                                           ~

BYtJ V56609 02'9 APPLICABLE DESIGN DOCUMENT (S) R1 N/A R2 SAR SECTION(S) UNID SYSTEM (S) R3 N/A A R1 R2 R3 Safety related? Yes No

     ~ Revision e EGN No. (or indicate Not Apphcable)

N/A Compute the Seismic Margin c Td~ jj High-Confidence-Of-tow-Probability-of y Failure (HCLPF) acceleration capacity Rev ewed F of the Auxiliary Control Building in

         /1//ff /M                                                                           _

accordance with the Conservative Approved ~ Detenninate Failure Margin (CDFM) f approach.

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Gl7l91 i g. FList'al1 pages added I term by this revit, ion M j List all pages deleted by this revision spu. List all pagos changed v4"*i by this revir. ion These calculations contain unvenfied assumption (s) Calculation contains special requirements or l that must be verified lator. Yes ]No limiting conditions. Yes ]No ] l l Abstract r The calculation documents the seismic margin assessment of the Auxiliary Control l Building in support of the SQN IPEEE. The Auxiliary Control Building is shown to l l exhibit seismic capacity in excess of the Review Level Earthquake (RLE).

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e L' REVISION LOG w_sg IPEEE SEISMIC MARGIN EVALUAT10N ;cc gy cc35 44 g

Title:

AUXILIARY CONTROL BUILDING

  ~'                                                                                  Date Revision DESCRIPTION OF REVISION                       Approved :

No. , O Initial Issue . . The number of pages in this calculation is 52 . Legibility evaluated and accepted for issue. l This applies for every page.

                       /      /DO                &f,9lY                             ly Signed                      Date/
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           '                                                                                               Page 3l21 CALCULATION DESIGN VERIFICATION (INDEPENDENT REVIEW) FORM sce.5H-oo46           h q                 rer eu nn >e        *W                          0
 ;                                                          Revision Calculation No.

Method of design verification (independent review) used (check method used): X

1. Design Review

2. Alternate Calculation l

3. Qualification Test Comments:

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! . ~ . - ' ' SHEET NO. 4 /d JOB NO. S'L'Id .0< JOB"T+ &- SC U % W BY h DATE G-M-% CALC. NO. SUBJECT h CHK'D PN DATE (o-29 W G.5 sVf-oo6 TABLE OF CONTENTS Subiect Eggg C O V E R S H E ET. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 TA B L E O F R EV1 S t O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 CALC U LATIO N D E SIG N R EVI EW .. . .... .. . ....... . ........ .. . . ... . . . .. . .. ..... .... 3 TA B L E O F C O NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

1. PURPOS E................................................................................... 5

2.

SUMMARY

.................................................................................5

3. TE C H N I C A L A PPR O A C H . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

4. REFERENCES..............................................................................7 5.

 }                                                   C A L C U LAT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 Comparison of Peak Response Accelerations ................................. 8 Scaled S S E Shear Wall Stre sse s ... ......... ....... ................... ........... .. 8 i                                                     investigation into SSE Shear Stresses .........................................10 Approximate Shear Wall HCLPF Capacity ....................................13 l                                                     High Bay R oof Dia phragm Evaluation...........................................15 l

l 1 l ATTACHMENT A - EXCERPTS FROM REFERENCE 3 ) ATTACHMENT B - EXCERPTS FROM REFERENCE 4 l l bTPr-CA VCA.)T L - D C.'iiKT5 M AWDWWCE 9 ) i i l 4

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SHEET NO. TlO JOB NO StW K JOB W &- CCt O T MA BY h DATE le-1A-% - CALC. NO. SUBJECT At % CHK'D N DATE b* DM U ~M ' # 1. PURPOSE The purpose of this calculation is to screen the auxiliary control building (Reference

1) at the Sequoyah Nuclear Plant (SON) Units 1 and 2. The results of this screening evaluation are used in the seismic margin assessment (SMA) of SON for resolution of the seismic Individual Plant Examination of External Events.

2.

SUMMARY

The high bay roof diaphragm has the lowest seismic capacity of any auxiliary control building structural component. The diaphragm was evaluated following criteria for seismic margin assessment intended to ensure that attached equipment will remain functional, even though there are no essential equipment mounted to the roof. The roof can damage essential equipment only if it collapses. The high bay roof diaphragm has a HCLPF capacity of at least 0.30g, and its collapse . f capacity should be even greater. It is concluded that the auxiliary control building  ! can be screened out.  ! j l i .i e i 2HD 289/sqnacb

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                   - . EQE INTERNATIONAL                                                                                       )
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SHEET NO. 6M i JOB NO. ENM JOB W 4-ECM su A BY M DATE G-7 b'k CALC. NO. SUBJECT Mk. CHK'D N DATE Y-

                        ""O                3. TECHNICAL APPROACH AND RESULTS STRUCTURE SCREENING Screening guidelines are recommended in EPRI NP-6041-SL, Revision 1 (Reference 2). Civil structures may be screened from detailed seismic evaluation for review level earthquakes with peak spectral accelerations less than 0.8g. The auxiliary                        i control building satisfies this criterion.                                                              .

Drawing Review: A review of the structural drawings (Reference 1)is performed. The only potentially significant seismic vulnerability identified is the shear capacity of the high bay roof diaphragm at its connection to the walls on Column Lines t and y. Comparison of Peak Structure Response Accelerations: Peak structure response accelerations due to the review level and safe shutdown (design basis) earthquake , are compared in the following calculations. The accelerations due to the safe 4, shutdown (design basis) earthquake (Reference 3) cx;;;" those due to review level earthquake (Reference 4),by fecters of 2 er greeter. ., ey it j 6 h ~ d,ul , /.

      ,a               Approximate Shear WallHCLPF Capacity: An approximate HCLPF capacity for the J               shear walls is estimated. Shear wall shear stress due to the review level earthquake is scaled from the maximum stress reported for the SSE. The ultimate shear stress is calculated following EPRI NP-6041-SL recommendations, including several conservatisms. Based on this evaluation, an approximate HCLPF capacity of about 04' 0 is estimated.

O.Rg , High Bay Roof Diaphragm Evaluation: The shear-friction capacity of the high bay roof diaphragm at its connection to the North-South walls is evaluated following criteria in EPRI NP-6041-SL. The diaphragm capacity is found to be at least 0.30g following these criteria, which are based upon ensuring the functionality of attached equipment, and not collapse. 1 l l

      .  ..1
        -1 4

2Ho 289/sqnacb

5 EQE INTERNAflONAL

             }                                                                                                        l SHEET NO. ~l/t'7 JOB NO. S"L"LotW JOB Th - %0 6MMA-                                             BY D         DATE ti-M-%

CALC. NO. SUBJECT % CHK'D W DATE b'2A ~#I H sc/,-5H- oo46

4. REFERENCES

1. Structural drawings. Series Nos. 41N300 to 41N302,41N304,41N306, 41 N307, 41 N309 to 41 N 316, 41 N318, 41 N319, 41 N321, 41 N322,  ;

41N325 to 41N330,41N332 to 41N334,41N337,41N338,41N340, 41 N 342, 41 N344, 41 N346 to 41 N349, 41 N351, 41 N 355, 41 N356, 41N358,41N359,41N361,41N365,41N566,41N368,41N379,41N372, 41N373,41N376,41N377,41N381,41N383,41N385,41N388,41N393 to 41N395,41N397,41N398,41N470 to 41N482,48N700,48N701.

2. "A Methodology for Assessment of Nuclear Power Plant Seismic Margin i (Revision 1)," Electric Power Research Institute, EPRI NP-6041-SL, Revision 1 1, August 1991.

3. Tennessee Valley Authority, " Dynamic Earthquake Analysis of the Auxiliary Control Building and Response Spectra For Attached Equipment," Report No.

CEB-80-20-C, Revision 3.

4. EQE Engineering Consultants, " Auxiliary Control Building (AGB), IPEEE In-Structure Response Spectra," Calculation No. SCG-5M-0005, Revision O.

 .            5.      Final Safety Analysis Report, Sequoyah Nuclear Plant Units 1 and 2.

6. " Building Code Requirements for Reinforced Concrete (ACl 318-63),"

American Concrete Institute,1963.

7. " Code Requirements for Nuclear Safety Related Concrete Structures (ACl 349-90) and Commentary - ACI 349R-90," American Concrete Institute, 1990.

8. Young, W.C., Roark's Formulas for Stress & Strain, Sixth Edition, McGraw-Hill Book Company,1989.

9. Tennessee Valley Authority, " Earthquake Analysis of Auxiliary and Control Bldg.," SCG-1-100.

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r 1 1 EGE l EQE MERNATIONAL  ! l _ SHEET NO. 8/d JOB NO. SC41.06 JOB h - So R SN BY h DATE 6-4-% j CALC. NO. SUBJECT KY- CHK'D & DATE b-7M/V I tGCC-5Had 6 l COMPARISON OF PEAK RESPONSE ACCELERATIONS i l l l A comparison of the SSE and SMA horizontal response accelerations willbe performed. N-S and E-W OBE response accelerations are available from Figures A.1 and A.5 of Reference 3 (see Attachment A). SSE responses are two times the OBE responses. SMA responses are available from Reference 4 (see Attachment B). Peak response accelerations at the center of mass will be compared, since these are a measure of the direct shear forces on the structure shear walls. N-S DIRECTION Floor OBE SSE SMA SSE/SMA l l EL 688.5' O.095g 0.19g 0.21g 0.90 i EL 713.5' O.125g 0.25g 0.24g 1.04 l EL 732.5' O.1459 0.29g 0.27g 1.07 l EL 748.5' O.16g 0.32g 0.30g 1.07 EL 762.5' O.18g 0.36g 0.32g 1.13 EL 791.25' O.21 g 0.42g 0.48g 0.88 l l E-W DIRECTION l Floor OBE SSE SMA SSE/SMA EL 688.5' O.095g 0.19g 0.24g 0.79 I EL 713.5' O.13g 0.26g 0.31g 0.84 1 EL 732.5' O.16g 0.32g 0.38g 0.84 EL 748.5' O.18g 0.36g 0.42g 0.86 EL 762.5' O.19g 0.38g 0.47g 0.81 EL 791.25' O.215g 0.43g 0.72g 0.60 l

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CALC. Nh UBJECT 4 CIA CHK'D N DATE 6'2 M I Ratios of SSE to SMA peak response accelerations range from 0.88 to 1.13 in the N-S direction and 0.60 to 0.86 in the E-W directions. The differences in ratios I between the two directions are due to difference.s in the the SMA accelerations, since the SSE accelerations at a given floor are about the same.

                                                                                                                    ]

SMA accelerations are greater in the E-W direction because E-W response is dominated by a single mode with a frequency of about 8 Hz. Response in the N-S directions results from two modes: A coupled translation and torsion mode with a frequency of about 6 Hz and a translational mode with a frequency of about 10 Hz. ' SMA accelerations appear to increase significantly at EL 791.25. The E-W acceleration appears to be influenced by a higher mode. The N-S acceleration may be influenced by torsion of the upper parts of the structure. I l

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TV A 10752 (CE SGI L. RBa% CEB REP. ORT " ' ' - REPORT NO. i l DYNAMIC- LAf,i'!D(IE,: ANALYSIS OF THE AUXILIARY CONTROL CES-80-20-C

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TABLE I ELEMENT PROPERTIES E = 590,400 K/Ft ;G = 236,160 K/Ft North-South Motion East-West Motion Torsion Moment.of Moment of Element Length. Area, Const t, Inertia, Shear Inertia, Shear No. Ft Ft2 Ft Ft k Factor Ftk Factor 1 T.62 13992 3020 x 10 5367 x 10 3.995. 6944 x 10 3.797 2 7.63 13992 3020 x 10 5367 x 10 3.995 69hh x 10 3.797 b 3 4.25 13992 3020 x 10 536T x 10 3.995 6944 x lo '3.797 h 8.38 11160 29ho x 10 5992 x 10 2.536 7518 x 10 3.323 5 8.37 11160 29h0 x 10 5992 x 10 2.536 7518 x 10 3.323 6 8.25 11160 29h0 x 10 5992 x 10 2.536 7518 x 10 3.323 h h b' T 9 25 8132 2510 x 10 4T11 x 10 2.011 7476 x 10 2.h75 h h h h 0 8 9 75 8132 2510 x 10 h711 x 10 2.011 7476 x 10 2.h75 1 9 16.00 7226 2920 x 10 hh92 x 10 2.107 6182 x 10 2.882 10 10.00 50kB 1810 x 10 2529 x 10 2.673 3507 x 10 2.203 11 k.00 50h8 1810 x 10 2529 x 10 2.673 3507 x 10 2.203 12 15.00 1795 55h x 10 789 x 10 - 1.777 1053 x 10 2.287 h h b 13 13.75 657 209 x 10 80 x 10 3.422 hh3 x 10 1.113 y

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87/07/30. I TENNESSEE VALLEY AUTHORITY l SEQUOYAH NUCLEAR PLANT I AUXILIARY CONTROL BUILDING EAST - WEST DIRECTION ! GROUND ACCELERATION = 0.09 G l SPECTRAL EARTHOUAKE RESPONSE l 5.0 PERCENT DAMPING HORIZONTAL ACCELERATION l m

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l SHEET NO. to/d JOB NO. StW D6 JOB Hk - 500 SW BY %!! A DATE (s-k -St CALC. NO. SUBJECT W CHK'D @ DATE b"M~N , 141-sH-oo44 SCALED SSE SHEAR WALL STRESSES Maximum shear stresses in the structure shear walls due to the SSE are reported in Section 3.8.4.5.1 of the FSAR (Reference 5). Approximate stresses due to the ' review level earthquake can be estimated by scaling these stresses by the inverse of the ratio SSE/SMA calculated on the preceding pages. Scaling based on peak response accelerations at the center of mass is reasonable since these response parameters reflect inertial forces on the structure. The following scale factors will be used: N-S Direction. A scale factor (1.0/0.88) = 1.14 will be conservatively used at all  ; elevations based upon the lowest SSE/SMA peak acceleration ratio of 0.88 at EL i 791.25'. E-W Direction: Above EL 763', a scale factor (1.0/0.S0) = 1.67 will be conservatively used based on the SSE/SMA peak acceleration ratio of 0.60 at EL 791.25'. Structure stresses at lower elevations should be much less sensitive to < the acceleration at EL 791.25', which has relatively low mass. Below, EL 763', a l scale factor of (1.0/0.79) = 1.27 will be conservatively used based on the lowest  ; SSE/SMA peak acceleration ratio below EL 763'. N-S DIRECTION Scale Approx. Story SSE Stress Factor SMA Stress EL 669' to EL 690' 146 psi 1.14 166 psi EL 690' to EL 714' 104 psi 1.14 119 psi EL 714' to EL 734' 98 psi 1.14 112 psi i l EL 734' to EL 749' 90 psi 1.14 103 psi l EL 749' to EL 763' 92 psi 1.14 105 psi j EL 763' to EL 778' 176 psi 1.14 201 psi Above EL 778' 234 psi 1.14 267 psi , l l l l .

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                 'M SHEET NO. u/M JOB NO. SEM Os JOB Th - SQR WW                                   BY h     DATE u-E  'i'f CALC. NO.                 SUBJECT Ath                            CHK'D    DATE b~M'N su-Sn - co n E-W DIRECTION Scale   Approx.

Story SSE Stress Factor SMA Stress EL 669' to EL 690' 112 psi 1.27 142 psi EL 6'90' to EL 714' 120 psi 1.27 152 psi EL 714' to EL 734' 90 psi 1.27 114 psi EL 734' to EL 749' 96 psi 1.27 122 psi EL 749' to EL 763' 62 psi 1.27 79 psi EL 763' to EL 778' 114 psi 1.67 190 psi Above EL 778' 60 psi 1.67 100 psi 1 1

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l TITLE . PLANT / UNIT Seismic Margin Assessment for the 480 V ERCW MCC SON 1&2

 ,   ,    ' PARING ORGANIZATION                  KEY NOUNS (Consult CCRlS LIST)             A///f - dp/g egj
 .   \         NE/EQE                                              Seismic, IPEEE, Margin, Category I, Equipment BRANCH / PROJECT IDENTIFIERS               Each time these calculations are issued. preparers must ensure that the original (RO) RIMS accession nurrter i is filled in.                                                                                                I

SCG-5M-0027 Rev (for RIM'S use) RIMS accession number RY9 '950609 01T APPLICABLE DESIGN DOCUMENT (S) R1 l NA l l\ R2 SAR SECTION(S) UNIO SYSTEM (S) R3 l NA 20 / -NA- l Rsvision 0 R1 R2 R3 Salety related? Yes No Q i

i EGN No. (or ind:cate rgApphcable) Statement of Problem ) I Prtpared

     'NN                                                                                                  CaWate me h.sm May l

Chrcked g , p//d {. HCLPF value for the 480 V ERCW  ; Rsviewed y MCC as part of an IPEEE review. l Adoer S-W  ; Appr 7k~ MdYNV 86~8 90

       ?* x sc h /9g-                                                                            G4 99-20 4 7 &                                               l

! k ,, List all pages added gg ggggf gg g form by this revision List all pages deletod pcv6SO $5'00 7 I by this revision spx. Ust all pages changed W by this revision l These calculations contain unverified assumption (s) Calculation contains special requirements or , that must be verified later. Yes No Q limiting conditions. Yes ] No Q Abstract The purpose of this calculation is to assess the seismic margin HCLPF capacyit for the 480 V ERCW MCC. This calculation employs the EPRI Margins Methodology (Ref M and a NUREG CR-0098 median spectral shape at the rock outcrop resulting in a free field ground motion ZPA of at least 0.3g, and is completed as part of an IPEEE review. Equipment

Description:

480 V ERCW MCC g g gg Located at Elevation 704 ft in the ERCW Building YC fr Grer */W G fck er e /Y hCWMfW WfW NN 3/ do u'W#a'c e . How ever,. tk pre sxd checker, were - nw & as .,./, 9,,pa,ei,n - ,s e a s , a n 1 doc m w & a t a ,, /,, ,, g,, , ,, ,,,, ,, ,, ye ,,,n=, \ e Nebolo WdC/ /toi ko/u/e- thu sdrcogg W\ p 6- &Mpr 4 O Microfilm and store calculations in RIMS Service Center. Microfilm and destroy. O N iiouv Liundi s.n A uc m c. O'L W Lb. b rv L'AL 1A

se.s - GM - oo 2 7 Page 2 //b REVISION LOG

        -Title. Seismic Margin Assessment for the 480V ERCW MCC's                           O Re sion 9,
                        .                 DESCRIPTION OF REVISION                               A    ved O              Initial Issue                                                     /fy'--

The number of pages in this calculation is 3I . Legibility evaluated and accepted for issue. This applies to every page Signed

                                                                       &/Y/97
                                                                         ' Ciated l                          -

l 1 l

   *'.         '-                                                                                                                                              Page 3/n.

CALCULATION DESIGN VERIFICATION (INDEPENDENT REVIEW) FORM i m - S M - co *?.n d , Calculation No. Revision l t Method of design verification (independent review) used (check method used):

1. Design Review

( l 2. Altemate Calculation ! 3. Qualification Test ] l l Comments: j l I t This calculation, including the computations, approach, judgements, and results has been found to be in accordance with the CDFM methodology presented in EPRI NP-6041-SL, ' Revision 1. The identification of the controlling failure mode and the resulting High-l Confidence-lew-Probability-of-Failure (HCLPF) seismic acceleration capacity appears to be consistent with the condition of the components observed during the seismic capability j walkdown. I N kev /rWCrlJ NdYef c4fcg/4 6'rn,

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AMCHWM , l l l' . l .l: , INFORMATION ON[7] I LYP)C74J') NEW FSAR DWG: .YES X NO 1 AFFECTED DRAWING R EV. CATEGORY DF.XA \l. N/A 3 NEW ISSUE (Mu GILL PLAN - ADoiiloN AL INFoEMATt04)\ SCAM : NTS UCEPT AS NOTEC (,"INPtCAL 4 ?LME4, B2./?W PUMP t96r 6TAThN ). PROJECT FACILITY UNIT 1 NcTW.: WoT'To BE. N corJef kTT O 04 DWCT: ERCW PUMPING STATION 3 7"" ' 7 41004 "TW5 9cA W tTH M 09 S SB-ca /, ' TITLE: ELECTRICAL EQUIPMENT 6.9KV SWGR, 480V MCC'S & XFMR'S P'_ANS, C.CMPA4 tow Pa?.AWIW6r: 1.t-3Cwl.31'2-SECTIONS & DETAILS PREREQUISITES: N/A OCA.M09858- 002 SEQUOYAH NUCdAR PLANT Q ANTICIPATED CCD: 1,2 35W3121 TENNESSEE VALLEY AUTHORITY

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                                                                                                                       '1 SECTION A - A                                          in 1NFORMkIngOMllj l

i NEW FSAR DWG: i YES _X_ NO NOTE: NOT TO BE INCORPORATE ON DWG.: AFFECTED DRAWING REV. CATEGORY WORK THIS DCA WITH M09858-o ol, .oot, - ou 4 Tsu -o o 7 NEW ISSUE N/A 3 COMPANION DRAWING: 1.2 35W312 SCALE : NTS EXCEPT AS NOTED ] PROJECT FACILITY UNIT ERCW PUMPING STATION 1 TITLE: ELECTRICAL EQUIPMENT 6.9KV CONTRACT NO. N/A SWGR. 480V MCC'S & XFMR'S PLANS, SECTIONS & DETAILS PREREQUISITES: N/A DCA-M09858- 003 SEQUOYAH NUCLEAR PLANT Q ANTICIPATED CCD: 1,2-35W312-1 TENNESSEE VALLEY AUTHORITY l

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l 0 _ tl h DCN M093S8-A REVISION RESP. ENGINEER LEAD ENGINEER DESIGN VERkFIED CHANGE REFERENCE

SON 1&2 PROJECT SHEET If OF A CALC. I.D. SCG4M00171 CALCULATION SHEEI hI Prepared by 5 06 Date 7/28/13 52 207 O 5 -C -Ob Z ] scG sH oo z. t Checked bM f,pff Date i g{cib aUBJECT: _ SEISMIC OUALIFICATION - REFERENCES THIS SHEET ADDED BY REV. 3

                            ' EVALUATION OF 3/8" 4 THREADED ROD (ANCHORAGE TO C3 SILL CHANNEL)

The 480V MCC unit is mounted on (2) C3 sill channels @ 704'-6' elevation in the ERCW Pumping Station. Use Elevation: 704'-6'. Damping: 2% , Condition: OBE , Damping 3% Condition SSE Building Response Spectra:(ref.-6) 2% OBE 3% OBE 3% SSE Direction Peak Acceleration Peak Acceleration 2 x 3% OBE North South 1.27 g 1.01 g 2.02 g East West 1.61 g 1.32 g 2.64 g Vertical 0.20 g 0.17 g 0.34 g Dead weight of MCC (5 bays) = 5 3250.00 lb. 7 ,y 7[f The 480V MCC is molded as a cantilever structure, which has a single degree of freedom.

               . the 1.5 multiolication factor for multi mode effect is not required. SSE peak acceleration values are used to qualify mounting bolts and welds.
     }

Seismic Loads: Direction OBE SSE North-South (Fx) 1.27 x 3250 = 4128 lb 2.02 x 3250 = 6565 lb East West (Fz) 1.61 x 3250 = 5233 lb 2.64 x 3250 = 8580 lb Vertical (Fy) 0.20 x 3250 = 650 lb 0.34 x 3250 = 1105 lb Loads at C3 sill channel for 3/8' rod evaluation: SSE load, loading direction (X, Y) SSE load, loading direction (Y, Z) Fx = 6565.00 lb. Fy = - 2145 / 4355 lb. l Fy = 2600 / -3900 lb. Fz = 8580.00 lb. Mz = 6565 x 45 = 295425.00in-Ib Mx = 8580 x 45 = 386100.00in-lb Loads at embedded plate for weld evaluation: SSE load, loading direction (X, Y) SSE load, loading direction (Y, Z) Fx = 6565.00 lb. Fy = - 2145 / 4355 lb. d 8580.00 lb. Fy = 2600 / -3900 lb. Fz =

  .f Mz = 6565 x 46.5 = 305273.00in-lb            Mx = 8580 x 46.5 = 398970.00in-lb

Page No. 05 i i , 52207.o 5 -C- cv7 f y_ m No. !*33504 l Scc,.cm. m cat.cD M 4 " *

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